o S e ORNL-TM-3039 Contract No.'WeTHOS-eng—26 Reactor Division MSRE SYSTEMS AND COMPONENTS PERFORMANCE by | Molten-Salt Reactor Experiment Staff Edited and Compiled by R. H. Guymon JUNE 1973 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. OAK RIDGE NATIONAL LABORATORY Oak Ridge, Tennessee 37830 — operated by UNION CARBIDE CORPORATION - . for the | U. S. ATOMIC ENERGY COMMISSION ‘ This report was prepared as an account of work . sponsored by the United States Government. Neither , the United States nor the United States Atomic Energy ! Commission, nor any of their employees, nor any of . their contractors, subcontractors, or their employees, { makes any warranty, express or implied, or assumes any i legal liability or tesponsibility for the accuracy, com- | pleteness or usefulness of any information, apparatus, . product or process disclosed, or represents that its use . would not infringe privately owned rights, . NOTICE ' 3‘% %SEER DISTRIBUTION OF THIS DOCUMENT 1S UNL\MI_TE%)\ o i . , 111 CONTENTS ABSTRACT « « oo v v v e e e e e e e e et et eeen e 1 1. INTRODUCTION .+ v v o o o o o o o « o v o s s o o o s o v vy 3 2. DESCRIPTION OF THE PLANT '+ « « v « « o o o o o s o s o o v+ & 3. CHRONOLOGY OF OPERATION AND MAINTENANCE . « « « o « s o s o« o 9 4. PLANT PERFORMANCE AND STATISTICS. « o o o o o o « o + o s o o 15 4.1 Cumulative StatisticS. o o « « o o o o o o o o o o o o 15 4.2 Availability during Various Periods. . 4« ¢« « &+ & « « 18 4.3 Interruptions of Operations. « « « o o o o o o ¢ o o o 22 4.4 Time Required for Operational Tasks. « o« o o o « o « o 48 4.5 Changes Made in the Plant. . . . . & e o s o s s o o 59 4,6 Tabulation of Recorded Variables at Full Power . . . . 63 5. FUEL SYSTEM &+ « ¢ o « o 5.0 o o o o o o« o s s o o o 5.0 o+ ¢ 89 5.1 Descriptione « o o o o o o ¢ o o o o o s 6 o o o o o » 89 5.2 Purging of Moisture and Oxygen from the System . . . . 89 5.3 Fuel-Circulating-System Volume Calibration , « « « . . 91 5.4 Drain TImeS. « o« o o o s o o o o o s s o s s o o o o o 91 - 5.5 Mixing of Fuel and Flush Salts . o 0 4+ & & « o « o« » 93 \=j 5.6 Primary System Leak. & ¢ ¢ o « s e o o % o e s 2 s o o+ 93 5.7 Operation '« « o ¢ o s o o 0 o 0 s 0 g 0 o 0 o 0 o s 94 . 5.8 Fuel Pump and Overflow Tank. « « « o o o o o ¢ « o o o 94 5.9 Primary Heat Exchanger « « « o« o o o o s o » ¢ o o o « 110 "5.10 Reactor Vessel . and Reactor Access Nozzle , o « ¢ + o o 112 - 5.11 Fuel and Flush Salt Drain' Tanks. ¢ « + ¢ o ¢ ¢ ¢ o =+ ¢ 129 6. COOLANT SYSTEM & &« o o « o o o o o o o s o s e.s ¢ o o o o '« 137 6.1 DeSCription « « o « o ¢ o o o 0 0 o o s s o o o o o o 137 6.2 Purging Moisture and Oxygen from the System. . . o . 137 6.3 Coolant Circulating System Calibration and Drain Time, 137 6.4 Operation....'........'...".oo...-q139 6.5 Coolant Salt Circulating PUmp. « « o o o .o ¢ o o ¢ o o 139 6.6 Radiator..............,.7....—...,.140 6.7 Performance of Main Blowers, MB-1 and MB«3 , , . ., o o 147 6.8 Coolant Drain Tank o« o o « « + .+ o o s 0 s + 0 .o ¢ ¢ ¢ 159 7. COVER-GAS SYSTEM ¢ o e e o oo o & o o s o s o s s o o o + o161 7.1 Initial Testing . .7 e ® e ® ® 8 o v s e e e s e e ¢ 161 7.2 Normal Operation e« + o o + @ « o s o o o o o o ¢ v o o 162 7.3 Conclusions and Recommendations. . « o o + « v o + o o 164 10. 11. 12, 13. 14, OO WOW\O W - 3 iv OFF-GAS SYSTEM L e ‘s & @ ; * e e o e e e e 8.1 ™ 00 o ™ nHWwN . 8.6 FUEL AND COOLANT PUMP LUBE OIL SYSTEMS. o o o « ¢ o » o - . vk WO WO WO\ - . =0 00N [l =] Description. . - | ] * L J - » & - . - o & - Experience with Coolant Salt Off-gas System. Experience with the Fuel 0Off-Gas System, Difficulties with the Fuel Off-gas System. . Subsequent Operating Experience with Fuel 0Off-gas System -* * - - * - @ * L - - * o ._ * . * Discussion and Conclusions « « o« o o o o o o Description. s J'.,. * % s e e e e * s » ¢ s Installation and Early Problems. . « « « « + & Addition of Syphon Tanks to the 0il Catch Tanks. . Oil Leakage. . ; e o @ ; 0 * & & = s & 8 09 0:0 . @ Change~out of O11 PumpS. o« « « o o o o o o o o o & " Test Check of One 0il System Supplying Both the Fuel - » * ® » . - - and Coolant Salt PumpsS « + « « o ¢ o o o = o s s o o 0il Temperature Problems . . « « ¢ o o o ¢ o o o o s o Packages. Increase in Radiation Levels at the Lube AnaIYSis Of 011. t.o » !-n * * & & & 2 Replacement of 011 . . ¢« & & ¢ o« ¢ « o & Discussion and Recommendations . « « o+ » COMPONENT COOLING SYSTEMS &« o o ¢ o s s o o s .0 10.1 10.2 Primary System . « « o« o« ¢ o .¢ ¢ s ¢ o & Secondary System ¢ ¢ o e s s oo ; . s s VENTILATION SYSTEM « « o o o o o o o o o s ‘o 11.1 11.2 11.3 Description. . + ¢« ¢ ¢« o o ¢ o ¢ ¢ o o o Operating Experience .. « « ¢« ¢« ¢« « « o & Conclusions and Recommendatfons. . « « o+ WATER SYSTEm » * * - . .. 2 - - L] - . - » . @ . - 12.1 - 12.2 12.3 12.4 12,5 12.6 12.7 LIQUID 13.1 13.2 13.3 Potable Water System . . . Process Water System , ., . Cooling Tower Water System Treated Water System . . . Condensate System . . « o o & Nuclear Instrument. Penetration. . General Water Systems Conclusions. WASTE SYSTEM &+ « o « o o s o o = o o ¢ * * e - * o & @ e o e & e Description. « « o« ¢ ¢ o ¢ ¢ o o s o o o Preliminary Testing.“. s s » {’o . s o e Operating Experience . « ¢« « ¢ s s o« o & HELIUM LEAK DETECTOR SYSTEM . » o« & ¢ o o s o & 14.1 14.2 14.3 14.4 Description and Method of Operation. . . Calibration . . . ¢ « ¢ o s o s ¢ « o Operating Experience . « « o« o » o o + Discussion and Recommendations . . . . . 0il v * e ® . 8 . - e 8 e e e & e @ 220 222 223 223 231 232 232 . 232 . 235 237 237 238 239 240 245 245, 246 - 247 247 . 247 248 251 « 251 - 251 - 255 256 15. 17. 18. "16.1 Description. « « + « o o o 16.2 Alternating Current System 16.3 -250—V dc System. . % e e » - 16.4 Reliable Power System. . . .. 16.5 48-V dc System . . + . . . 16.6 Diesel Generators. . . . . 16.7 100-kVa Variable Frequency 16.8 ConclusionS. . « « o « o & HEATERS , & ¢« ¢ & ¢ o o« o s ¢ o o - 17.1 Description. + « « « + . . 17.2 Preoperational Checkout., . ~17.3 System Heatup, . « + « + & ~17.4 Heater Performance . . . . -17.5 Systems Cooldown Rate. . . 17.6 Discussion . . ¢« « « & o o 'SAMPLERS. et e e eie 4w e we s 18.1 Fuel Sampler—Enricher. . o 18.2 Coolant Sampler. . « « + o 19. - 20. INSTRUMENT AIR SYSTEM . « o o « o 15.1 15.2 - 15.3 16, Description . . + « & + & Operating Experience . . . Conclusions, . « + o« & « & ELECTRICAL SYSTEM . e e 9 * & e CONTRDL RODS; *® » s 2 & & o = . % : 2 19.1 19.2 19.3 19.4 19.5 19.6 19.7 - -FREEZE 20.1 20.2 20.3 "7.20‘4_ 20.5 FREEZE 21.1 21.2 21.3 Description., « + « + & Initial Testing. . « + & Periodic Testing . . . Operating Experience , Improved Rod Scram Testing ‘Maintenance Experience . . . . - * . - .. Discussion and Recommendations VALVES . ; . .-.'.'; '7‘ . ; . IntrOduction é.j ;@c " s 4 s s & 8 s e the Freeze Valves of the Freeze ~ Description of the Design of Description of the Operation Operating Experience , . . Recommendations.-.-. o 6 FLANGES. .f;:.-.i!TF o“o ; Description. « « « + & « & Operation. ¢ o . ¢ 8 8 & u Conclusions. . + ¢ « o« ¢ & * - . - . - *» or—Generator . - . . . . - * * * . » o - . [ . - * & * . * . » - e e e e o » . » o ® e » e ® . * . L] * * - . * . ® - » . * ® . . * * - . - . - * e o * o = - * Valves. . . » . - - . e o — s i et vi 22. CONTAIMNT * - » - . » - ) . * - o & - . * - . * - * . * ® . - 22.1 22.2 22.3 22.4 22.5 Description and Criteria .« . « o « o o o « o o o o o & Methods Used to Assure Adequate Containment and ResultSe « o s o o ¢« ¢ o ¢ o o o o o 5 o 5 o s s o » Discussion of Cell Leak Rate Determination Vapor Condensing System . . ¢« ¢ ¢« ¢ o « & Recommendations. . « « o« « ¢« ¢« ¢ o« ¢ o o & 23. BIOLOGICAL SHIELDING AND RADIATION LEVELS . o & « ¢ o ¢ o « & - 23.1 23.2 24. INSTRUMTATION .. ® [ . - * e . * @ * - * . . * ® » . 24.1 24.2 24.3 24.4 24.5 24.6 24.7 24.8 24.9 24.10 24.11 24.12 - 24.13 - 24.14 - 24.15 24.16 24.17 24.18 24.19 24.20 24.21 REFERENCES . Description- * & s 2 s & o s e . & @ Radiation Surveys — Approach to Power. « « o« « o« o o o Radiation Levels During Operation. . « « « s o s s & & ConclusionS.: « + « s o o ¢ o o s s o o o o ¢ o o o » o IntrOduction . - . » - » » » » » * e L Initial Checkout and Startup Tests . . Periodic Testing . « « ¢ ¢ ¢ s ¢ o « o« Performance of the Nuclear Safety Instrumentation. Performance of the Wide-~Range Counting Channels. . Performance of the Linear Power Chamnels . . « . . BFs Nuclear Instrumentation. . . + ¢« o + & ¢ o o o« Nuclear Instrument Penetration . . « « « « ¢ o + Performance of the Process Radiation Monitors. . . . Performance of the Personnel Radiation Monitoring and Building Evacuation System . « « « o ¢ o o ¢ ¢ o o o Performance of the Stack Monitoring System . . . . . . Performance of Thermocouples and the Temperature Readout and Control Instrumentation. . . . . . Performance of Pressure Detectors. . « « « « &« Performance of Level Indicators ., . . . . . . Performance of the Drain Tank Weighing Systems - Performance of the Coolant Salt Flowmeters Performance of Relays. . « « « . . . Training Simulation . . « « ¢ « & Miscellaneous ., « o« ¢ o ¢ & o o Conclusions and Recommendations. . - * * * . . . * . - . . - . » [ ] - . * L] » a - e » - * - . - . . & * * . -2 - > * . . . - - * - ® - s .9 - - . 403 403 403 404 404 404 408 408 409 410 411 411 412 412 413 MSRE SYSTEMS AND COMPONENTS PERFORMANCE ABSTRACT When the MSRE was shut down in December 1969, it had accumulated 13,172 full-power hours of operation. Salt had been circulated in the fuel system for 21 788 hours and in the coolant system for 26, 076 hours. _ Essentially no difficulty was encountered with the primary system during operation. After the reactor was shut down there was an indication of a leak in the drain-tank piping at/or near a freeze valve. Further in- vestigation will be maderlater as to the nature and cause of this leak. There was a small continuous 1eakage of lubricating oil into the fuel pump throughout the operation. This, together with salt. mist, caused peri- odic plugging in the off—gas system which was designed for clean helium, Filters installed in the main lines proved very effective., In early operation, difficulty was encountered with the coolant radi- ~ator. The doors would not seal, there were too many air leaks, and thermal insulation was inadequate. After these were repaired, the system operated fine except for a failure of one of the main blowers and some trouble with the blower bearings. , . , , Only relatively minor difficulties were encountered with the contain- ment and other systems. 1. INTRODUCTION P. N. Haubenreich Operation of the MSRE constituted a major step toward the objectives of the Molten-Salt‘Reactor Program. The g6a1 of this program is the de- - velopment of'large, fluid-fuel reactors having good neutron economy and producing low-cost electricity.1 The MSRE was built to demonstrate the practicality of the molten-salt reactor concept with emphasis on the compatibility of the materials (fluoride salts, graphite, ‘and container ralloy) the performance of key components, and the reliability and main- tainability of the plant. In the course of 5 years of testing and operation of the MSRE (196h - 1969) the operators sccumulated con51derable experience with the various components - and systems in the reactor plant. Tnis experience, properly disseminated, should be valuable in the continuing development of molten- salt reactors. Much has already been published in tne Molten Salt Reactor Proéram semiannual progress reports (Refs. 2 to 14) but such reporting is piecemeal and sometimes rathervcondensed. On the other hand, there is much very detailed information in test reports and operations and mainte- - nance files, but these are relatively inaccessible and specific informa- tion is tedious to extract. It was considered.worthWhile therefore, to extract, organize ‘evaluate and report the experience with MSRE systems and components. o The purpose of this report is to present a convenient, comprehen51ve description of the MSRE experience. The first chapters.briefly describe the plent end outline the chronology of its'Operation. Next there is a 3 chapter on the overall plant performance, 1nclud1ng statistics relative to reliability and maintainabillty. The chapters which follow are each devoted to one system or component. Finally, there‘is a chapter of dis-- cussion and conclusions. 2. DESCRIPTION OF THE PLANT R. H. Guymon The MSRE was a single—rigion;-circulating molten-salt'fueled‘ thermal reactor which produced heat at the rate of about 8 Mw. The fuel was UFy in a carrier salt of L1F-BeF2-ZrFq. At the operating temperature of 1200°F, this salt is a liquid which has very good physical propertles: viecosity ebout 8 cent1p01se den51ty about 135 Ib/ft3, and'vapor pressure less than 0.1 mm Hg. | | | ~ The design conditions are shown in the flow diagram (Fig. 2--1). The general arrangement of the'plant is shown in Fig. 2-2. The salt-containing ~piping and equipment was made of Hastelloy~N, a nickel—molyhdenumyiron- ‘chromium alloy with exceptional,resistance to corrosiOn'by molten fluorides and with high strength at hlgh temperature | In the reactor prlmary system, the fuel salt was . rec1rculated by the sump-type centrifugal pumphthrough the shell—and-tube heat exchanger and the reactor vessel. - The 5-ft diam. by78-ft high reactor vessel is shown in Figure 2—3‘ Tt was filled with 2-in. by 2-in. graphite moderator stringers which had grooves machlned in the sides to form flow channels ‘for the fuel , salt. ©Since the graphlte is compatlble with the.molten salt, 1t was possible to use unclad graphlte which is deslrable to obtaln good neutron economy. The heat generated in the fuelisalt as»it passedfthrough the reactor was transferred in the heat exchanger to & molten LiF-BeF; coolant salt. The "coolant salt was c1rculated by means of a second sump—type pump through : the heat exchanger and through the radiator. Air was blown by two axial flow blowers past the radlator tubes to remove the heat which was sent -:up the coolant stack where 1t was dlSSlpated to the atmosphere. _ / Drain tenks were prov1ded for storlng the fuel and coolant salts at hlgh temperature when the reactor was not operatlng., LiF-BeF, flush salt ~ used for flushing the fuel system'before and after maintenance was stored (\\jd in the .fuel flush tank. The salts were dralned by gravity. They were | ( transferred back to the circulatingrsystems by pressurizing the tanks with helium. STACK FAN U | R . el S g;mq 5 ","‘ | ol A coouant [ S | i ]enricher PUMP ; o i | { { J TO ABSOLUTE FlLTERs-.......i - ' | ‘ !- i 1018 *F ; i =i 850 G.PM. OFF-GAS AT R oow ' 1240°F _ — ' OVERFLOW TANK ABSOLUTE 170 *F FLTERS ' 120G GAM, BLDG. REACTOR : 1075 F VENTLATION . VESSEL | power- FREEZE FLANGE {TYR) . 8 Mw bl geress L OLAN FREEZE VALVE (TYP) o wmd COOLANT , F‘ SYSTEM 4 J 'h S8 -~ " RN A = reeresTosreme MAN CHARCOAL BED ORNL-DWG 65-14108 LEGEND ssmmmm FUEL SALT — COOLANT SALT stresreveneer mm CWER GAS ----- RADIGACTIVE OFF ~GAS : FH.TERS - - . DRAIN TANK ORNL-DWG 63-1209R REMOTE MAINTENANCE CONTROL ROOM - | i ' | REACTOR CONTROL ROOM _ ‘ 4 - . ‘l S 6 —' -" '. : ! _l'l i j | E; ‘ » —— » —_— AN i (,fi\ : J 9 8 1. REACTOR VESSEL 7. RADIATOR 2. HEAT EXCHANGER 8. COOLANT DRAIN TANK 3. FUEL PUMP 9, FANS 4. FREEZE FLANGE = 10. FUEL DRAIN TANKS 5. THERMAL SHIELD 1. FLUSH TANK 6. COOLANT PUMP 12, CONTAINMENT VESSEL 13. FREEZE VALVE Fig. 2.2 Layout of the MSRE ORNL-LR-DWG 61097R1A FLEXIBLE CONDUIT TO GRAPHITE SAMPLE ACCESS PORT CONTROL ROD DRIVES COOLING AIR LINES o ° __3_; o ‘ 1 ACCESS PORT COOLING JACKETS FUEL OUTLET r REACTOR ACCESS PORT " CORE ROD THIMBLES N i SMALL GRAPHITE SAMPLES ROD T! ==Y HOLD-DOWN ROD LARGE GRAPHITE SAMPLES 1 OUTLET STRAINER . & CORE CENTERING GRID FLOW DISTRIBUTOR VOLUTE GRAPHITE ~MODERATOR . STRINGER ’ Ty AL FueL iNLeT < (| s T~ CORE WALL COOLING ANNULUS REACTOR CORE Can —i| REACTOR VESSEL ANTI-SWIRL VANES — ' S ' e MODERATOR VESSEL DRAIN LINE - SUPPORT GRID Fig. 2.3 Details of the MSRE Core and Reactor Vessel The fission product gases, krypton and xenon, were removed continfiously_ from the circulating fuel salt by spraying salt at a rate of 50 gpm into the cover gas above the liquid in the fuel pump tank. There they trans- ferredifrom the liquid to the gas phase and were swept out of the tank by a small purge. of helium. After a delay of about 1-1/2 hr in the piping, this gas passed through water-cooled beds of activated charcoal. The krypton and xenon were delayed until all but the 8%Kr decayed and then were diluted with air and discharged to the atmosphere. | | The fuel and coolant systems were provided with equipment for taking samples of the molten salt while the reactor was operating at power. The fuel sampler was also used for adding small amounts of fuel to the reactor while at power to compensate for burnup The negative temperature coefficient of react1v1ty of the fuel and graphite moderator made nuclear control of the system very simple. However, three control rods were provided for adjusting temperature, compensating . for buildup of fission products, and for shutdown. The plant was provided with a simple processing facility for treating full 75~ft? batches of fuel salt with hydrogen fluoride or fluorine gases. The hydrogen fluoride treatment was for removing oxide contamination from | the salt as Hy0. The fluorine treatment utilized the fluoride vblatility process for removing the uranium as UF5.- Auxiliary systems included: (1) a cover-gas system with treating stations for remofiing dxygen and moisture from the helium cover gas; (2) two closed-loop oil-systems for cooling the fuel and.coolant pumps and providing lubrication to’tfié’bearings; (3) = closed 100p.component coolant system for cooling the: control rods and other in-cell components; (L) several cooling water systems including a closed-loop treated water system for cooling certain in-cell équipment§ (5) a ventiletion system for contamination control; andr(6) an instrument air system. ‘ A1l of the primary salf'éystém'was located in the reactor and drain 'tank cells. These sealed pressure vessels prov1ded secondary contalnment. For a fuller descrlptlon of the plant, see Reference 15 3. CHRONOLOGY OF OPERATTON AND MAINTENANCE - ' . R. H. Guymon | Design of the MSRE began .in the summer of 1960 and by August 196h installation was far enough along to permit the planned non-nuclear testing16 to begin. Milestones that were passed in the years which fol- lowed are listed in Table 3-1. Teble 3-1. Milestones in MSRE Operation Salt first loaded into tanks | October 2k, 196k Salt first circulated through core. January 12, 1965 First crltlcallty (23%5y) . June 1, 1965 First operation in megawatt range January 24, 1966 Full power reached | - May 23, 1966 Nuclear Operatlon w1th 235U concluded ) - March 26; 1968$ Strip uranium from fuel salt August 23-29, 1968 First crltlcal with 233U p._ October 2, 1968 Reach full power with 233y | ~ January 28, 1969 Nuclear operation concluded - Decenber 12, 1969 The activities durlng the perlod of non—nuclear testlng, zero—power - experlments, and preparation for power 0perat10n are outllned in Fig. 3-1. During the prenuclear testlng, except for the usual startup troubles due to early fallures and instrument malfunctlons, 21l systems operated - well and there vere no unant1C1pated problems in handling the molten _salt. Some plugglng of the offgas system valves and fllters was encoun- ~ tered, but thls did not seem serlous.__ Crltlcallty was attalned‘by addlng UFu~L1F enrlchlng salt to the '~eafr1er salt. More uran;um{was added to bring the fuel gradually to the . desired operating concentratien'while the control rods were calibrated and .feactivity:coeffiéients were measured. At the conclusion of'thefzero- power nuclear experiments‘the.reactor was shut down to finish construction of the vapor-condensing system, to carry out some inspection, and to INSPECTION AND 'PRELIMINARY TESTING MSRE ACTIVITIES JULY 1964 — DECEMBER 1965 PRENUCLEAR TESTS OF COM PLEATE SYSTEM FINAL PREPARATIONS " FOR POWERLOPERATION ” N FINISH LEAKTEST, - INSTALLATION PURGE & HEAT OF SALT SYSTEMS SALT SYSTEMS TEST AUX. SYSTEMS . OPERATOR TRAINING LOAD SALT TEST TRANSFER, OPERATOR. FILL & DRAIN OPS. TRAINING — ’ Al INSTALL INSTALL CORE SAMPLES CONTROL RODS ‘ INSPECT FUEL PUMP SAMPLER-ENRICHER HEAT-TREAT CORE VESSEL 277227 _ ; TEST SECONDARY ' TAI FINISH VAPOR-COND. SYSTEM CT MODIFY CELL PENETRATIONS ADJUST & MODIFY REPLACE RADIATOR DOORS RADIATOR ENCLOSURE : LOAD U-235 - INTO DRAIN TANKS LOAD & IN ZERO-POWER - _ LOW-POWER COOLANT FLUSH CIRCULATE CIRCULATE . NUCLEAR , ~ (0-50 kw) SALT SALT C & FL- SALTS CARRIER EXPERIMENTS | EXPTS. 2222772 g L } 1 { 4 | y | 1 { 1 1 § { . _ I I 1 RN 1 t I T b I 1 1 I 1 i I 1 1 J A S 0 N D J F M A M J J A S 0 N D 1964 1965 Fig. 3.1 MSRE Activities July 1964 — December 1965 - 0T 11 meke some repairs and modifications, including replacement of the heat- warped radistor doors. rThe'firstfteSt,of the secondary containment was made during this shutdown period After tests in the kilcwatt range showed that the dynamics of the system were as expected,,the approach to full power was started in January 1966. Plugging in the fuel offgas system occurred immediately after increasing the power tc one megewatt. Almost three mouths were spent investigating and remedying the offgas prdblem-by'installation of ‘8 large, efficient filter;'”As:shown in Fig. 3-2, the power ascension was ‘resumed in April and full power, which was limited by the capsbility of the,heat-removal system, wasjsttained in May. High-pcwer,operation'wes'abruptly halted in July when the hub and blades of one of the main.blcwers?in.the heat-removal system broke up. While the reactor'was down, 'the arrsy of graphite and metal”specimens'was removed from the core and & new array instelled. During the flushing operations before the specimen removal the erl-pump bowl was accidentally over-filled and some flush selt froze in the attached gas lines. ~ Tempo- rary heaters were installed remotely to clear the lines and in November the offgas line at the pump bowl was cleared by running a tool through it. Most of the remaining operation with 235y was at power. Restrictions continued to develop in the offgas system.but caused llttle delay in the program. Shutdcwns were necessary to make repairs on the fuel sampler- enricher, the component coolant pumps, and in-cell air line,disconnects o whose leakage caused an indication of high cell leak rate. The core r,:,specimens were replaced periodically, annual containment tests were made, '_ and verious experiments conducted. Operation with 23°y fuel was termi- nated on March 26, 1968 8o that the urenium could be stripped from the fuel selt and replaced with 23311 During the shutdcwn which fbllowed the core speclmens were replaced and the fuel piping and. vessels were surveyed using remote gamma ray spec- r;troscOpy to determine the distribution of fission products. Necessary and - pipreventive maintenance was also done in preparation for operatlon w1th -:-i-zsaU fuel. | | | After testing end extensive modification of the excess fluorine dis- ' posal system, the flush salt and fuel salt were processed to strip the DYNAMIKCS TESTS - INVESTIGATE OFFGAS PLUGGING REPLACE WALVES AND FILTERS | RaISE POWER } REPWR SAMPLER | ATTAN FULL POWER } ek conTammenT FULL~POWER RUN -— MAN BLOWER FALURE REPLACE CORE SAMPLES TEST CONTAINMENT AuUsH ] 12 FUEL SALT N FUEL LOOP POWER 0 RN FLusH ] 2 4 668 10 POWER (Mw) ° ORNL-0WG 69~ T293R2 HIGH-POWER OPERATION TO MEASURE & fag INVESTIGATE COVER GAS, XENON, AND FISSION PRODUCT BEHAVIOR ADD PLUTONIUM RRADIATE ENCAPSULATED U MAP F.P. DEPOSITION WITH GAMMA SPECTROMETER MEASURE TRITIUM, SAMPLE FUEL REMOVE CORE ARRAY PUT REACTOR IN STANDBY Fig. 3.2 Outline of the Four Years of MSRE Power OperatAion 13 ursnium and remove théfcorroeion products produced during the fluorina- tion.!? attached to a drain tank began immediately thereafter. The cells were closed and leak-testednbefore the reactor was made crltlcal by small ad- ditions of 233y through the sampler-enrlcher. ‘On October 8, 1968, Glenn Seaborg manipulated the controls ‘to raise 'the power to 100 kw maklng - 233 S U. Loading of 233U into the carrier salt through spec1a1 equlpment the MSRE the world's first reactor to operate on Early in the zero-power phy31cs experlments, the amount of éas en- trained in the;circulatingbfuel had increased from <0.1 to ~ 0.5 vol %. Investigation of this, as well'as dgifficulties with the sampler-enricher end plugglng in the offgas system, delayed the approach to full power | until Jenuary 1969. - | | | As the power was 1ncreased into the megawatt range, ‘there were ob- served for the first time sporadic small increases (~ 5 to 10%) in nuclear power for a few seconds, occurring with & varying frequency somewhere ' ~around 10/hr. The characteristics of the transients pointed to changes - in gas volume inh the fuel loop and it appeared that they were most 1ihely" caused by occasional release of some gas that collected 1n the core.wln This hypothesis was -supported when, late in February, a varlable frequency generator was used to operate the fuel pump &t reduced speed. The gas ent_rainment in the'ruel circul_atfng loop decreased sharply (from 0.7 to ?O.l vol %)-and the perturbations ceased entirely.’ Operations continued at various nuclear powers and. fuel-pump speeds until Junerlnwhen the reactormfias shut down to replace the core specimens, investigate the'distribution:offifission products-in the-prinary loop by ,gammafscanning, remove the'offgas restrictione and-test the secondary containment. For the first tlme one of the control rods dld not scram h‘l,and it was replaced durlng the shutdown. | e o 77; The remaining months of operet1on were spent in various studles of ’{1the behavior of tritium, xenon, and certain other fission products .in. ther 'reactor.:fln Novemberflt became ev1dent that . sufflclent funds would-.not 7 be“available_to'continne operation and on December 12, 1969 nuclear’ ' operation wes concluded. Another flrst occurred shortly after the fuel was drained'when a small amount of flssion—product activity appeared in 14 the cell atmosphere, indicating a leak in the primary system apparently near a freeze valve. , | | | | After the final shutdowfi, fihe facility was_placed_in a standby.con— dition to await post-operation examinatiohs plahned for early in the next fiscal year.19 The operating crewsrwere disbanded and overrthe next six - months most of the engineers were reassigned as they completed analyses and reporting of the reactor expefience. As of this writing, the post- operation examination has not been accompiished. - In this report and elsewhere MSRE run numbers are often used to identify the period of operation. Generally a new run number was assigned at the end of a major shutdown or when there.were substantial changes in . the purpose or type df operation. The starting dates for eaéh of the 20 runs are listed in Table 3-2,. Table 3-2. Dates 6f MSRE Runs Run No. ' Starting Date® Run No. ' Starting Date® 1 January 9, 1965 11 January 24, 1967 2 Msy 11, 1965 12 June 8, 1967 3 May 31, 1965 13 September 3, 1967 L December 5, 1965 1k September 19, 1967 5 February T, 1966 15 August 1k, 1968 6 March 26, 1966 16 December 10, 1968 T June 11, 1966 | 17 January 10, 1969 8 September 1k, 1966 18 April 11, 1969 9 November 6, 1966 19 July 31, 1969 10 December 6, 1966 - 20 ' November 20, 1969 Slhese are the dates on which samples, logs, etc., began to receive new numbers, and are not generally the dates of reactor startup. 15 4. PLANT PERFORMANCE AND STATISTICS R. H. Guymon ~P. N. Haubenreich "They 've kept the darned thing running and when they shut down they can get back on the line; you can't knock that!" An unidentified "AEC spokesman', ag - _quoted in October 12, 1967 Nucleonics Week. "So far the Molten Salt Reactor Emperzment has operated successfully and has earned a reputation for relia- . bility." USAEC Chairman Glenn T. Seaborg at the ‘ceremony marking the first operatton of a reactor fueled'wtth 233y, October 8, 1968. The MSRE ran long and it ran well. It ran long enough with 235U ., ~ fuel to attaan the original goals of the experlment then continued through more than a year of added experlments with 233U fuel. These statements are supported by the statlstlcs on the overall plant performr ange-presented in this chapter.. Later chapters describe the performance of individual components and systems which made up-the plant. 4.1 Cumulative Statistics Several different\ipdiggtions of how long the MSRE ran ére’presented in Teble L-1. The time critical is a commonly used index for reactor ex- - periments. Another basis of cOmfiarison With“othér reactors is the number of equivalent full-power hpurs; _Integrated power (Mw-hrs) is closely re- , léted. (The figures quoted'fiéféfaré'based on 8 full power of 8.0 Mw, - which is the value 1nd1cated by heat balances.20 Analyses of isotopic -changes 1nd1cate that full pcwer was 7.34 Mv, 1n which case the tabu- -_lated Mw—hrs should be mnltlplled.by 0.92.). The amounts of tlme that o salt circuleted in the loops are of 1nterest from the standp01nt of the 'demonstrat1on of materlals compatlbility.- Each pump operated: a length of ' ':tlme equal to the salt circulat1on plus helium circulation in that loop. . The number of thermal cycles of various kinds on dlfferent parts of the salt system are llsted 1n Table 4-2, The numbers are small in com- p&rlson with the perm1581ble numbers of cycles except in the case of the 16 Table 4-1. Accumulated Operating Data Time Critical (hrs) Integrated Power (Mw-hrs)® Equivalent Full-Power Hours Salt Circulating Time (hrs) Fuel Loop Coolant Loop Helium Circulating Time (hrs) ~ Fuel Loop | Coolant Loop Time Above 900°F (hrs) Fuel Loqp | Coolant Loop Fill and Dreain Cycles Fuel Loop Coolant Loop 235y operation 23%U Operation Total 11,515 6,140 17,655 72,401 33,296 105,737 9,005 4,167 13,172 15,042 6,746 21,788 16,906 9,170 26,026 4,046 3,384 7,430 3,172 1,535 4,707 20,789 10,059 30,848 17,44Y 9,99k 27,438 | 37 14 51 13 6 19 BBased dn heat balances which indicated full power was 8.0 Mw. Table 4,2 MSRE Cumulative Cycle History (. ‘Thaw Usage _ ' ! | | : Quench ‘& | Factor Component, - 'Heat/Cool |Fill/Drain|Power|Quench , Time . On/Off| Thaw: Trans. & - *These figfires are based on the original éalculations} If they were based on freeze flangé thermal cycle tests, the usage factors would be 23.04% and 13.37%. ' LT . 18 | freeze flanges. Here the fuel flanges experienced 99% of the permissible number of cycles based on early calculations or 23% ‘of the number perm1351- ble on the basis of extended tests of & prototype flange. In order to put the MSRE record in proper perspective, 1t is neces- sary to measure it against those of other reactors in a similar stage of development. A comparison on the basis of equlvalent full-power hours is made in Table 4-3. It is not invidious to say that the MSRE compares well with other reactors having the same purpose that is, to demonstrate the .practlcality of a reactor concept. 4.2 Availability during Various Periods The best index of the rellabillty of a plant should be the fractlon of time (over some extended period) that it is available for its intended service. Perhaps inevitably, however, in experimental,plants where the . obJectives include.more:than simply_generating‘pcwer, the definition of -availability'is not always s0 clearcut. Therefore we heve presented in Table U-4 an index which, in principle, isfiless significantfibut-whose' A definition is quite clear; namely,.the time that the réactor wes actually critical in-each 3-month period during the 4 years of power operation. Percentages of elapsed time are shown for selected intervals, but these must not be regarded as e measure of reliebility since the reactor was suberitical much of the tlme because the planned program requlred it. ',(Durlng shutdowns for core specimen removal or the substitution of tzaau for example.) ~ In the MSRE test progra.m16 it was planned that there be a perlod of ’ 0peratlon for the prlmary purpose of demonstratlng plant rellablllty. _iThlS phase of the program covered the last 15 months of operation with 235y, Table 4-5 gives a'breakdown of the tlme durlng thls period. The reactor was critical 80% of the time and the . avallablllty of the plant, as defined in this table, was 86% This amply Justifled Dr. Sesgborg' s | statement quoted at the head of this chapter. ‘ Another indication of rellabllity is how long a plant can be kept continuously on line. In the final run with zasU, the fuel salt was in the loop continuously for just over 6 months, before it was draimed for “ 19 Teble 4.3 - Equivalent Full-Power Hours® Produced by Early U. S. Reactors ‘of Several Types - Operation e w First Interval L Type Reactor Critical Terminated = {years) -~ EFPH Aqueous - _ Homogeneous HRE-2 12/57 5/61 3.4 3,100 - Organlc-Cooled OMRE 9/57 . 4/63 5.6 5,934 " Piqua 6/63 " 1/66 2.6 5,642 Sodium-Graphite 'SRE W57 2/6h 6.8 8,1h0 co . Hallam 8/62 9/6k4 2.1 2 661 HTGR Peach Bottom 3/66 ' ’ib 3.8 10,836° LMFBR EBR-1 8/51 | 'ié/62 11.3 5 sohb EBR-2 11/63 b 6.1 11,713, Fermi 8/63 b 6.3 102 PWR Shippingport . o (Core #1) 12/57 2/6k 6.2 27,781 BWR EBWR 12/56 6/67 10.5 11,164 VBWR 8/57 12/63 6.3 11,814 MSRE 6/65 12/69 4.5 13,172 MSR ®Calculated from.thermal'MW—hrs and installed capaoities reported in USAEC-DRDT booklet "Operating History of U. S. Nuclear Power Reactors — "'1969" except Shlppingport Core—l data from April 1964 Nucleonics. bOperation is not yet termlnated. EFPH quoted is through 1969, Total for Shipplngport (with two cores) through 1969 is h3 400 EFPH. 20 Table 4-4 Time Critical Each Quarter during the Four Years of Power Operation (1966 - 1969) Hours ‘ - Critical Time - Year - Quarter = Critical Elapsed Time 62 v | ) 1070 413 1221 1066 & w o 1852 | ’ T4 ' 235 1186 73.9% 1292 (1967) | « 9T.2% 21k (Qtr.) 1967 = w N R 20L45 0 0 735 1968 Eow o e _,i 1800 1375 1054 1176 1969 61.5% 56% - (1969) (233y) = w e 21 - - Tdble 45 Breskdown of Time during Sustained Operation Phase: =~ Of MSRE Program (December 1k, 1966 - March 26, 1968) Activity or Condition - - - Time - 'sePercentsge Criticel - 8934 nr 79.6 Available "~ Changing specimens - o 26d 5.6 086.3 - 'Annual tests - 54 1. 1 _Air-line disconnects (1/67) 13 d 2.8 | Sampler latch (8/67) | 35 d 7.5 Maintenance Component cooling pump (9/67T) 3d 0.6 1%? Sempler wiring (12/67)-_: ... -.34a - 0.6} e r View in. reactor cell (6/67) 6.4 1.3 Q"ther__-. 1 Mlscells.neous o Lo oo 0.9 2.2 'I'otal. ela.psed o S | | k68 4 ‘100.0_ | 22 | the planned processing. During these 188 days the reactor was counted as unavailsble only 63 h (1.4% of the time) while repairs were being made to the fuel-sampler drive. It was actually critical for 97.8% of the time. The fingl phase of the MSRE operation, with 233y fuel, was not aimed Primarily at demonstrating reliability,21 but it turned out that the availa- bility of the plant was still remerkably high. This is shown in Teble 4-6. Here available time includes (in addition to the critical time) subcriti- cal intervals during the zero-power experiments, time spent in changing the pump power supply during varidble—épeed;tests, changing the core speci- mens, gamma-scanning the drained salt systems, and’ experiments on behavior of gas in flush salt. | 4.3 'Interrfiptions of Operations FiSUre 4-1 provides a broad view of the MSRE operation and major ifiter— ruptions. This figure was prepared for inclusion (together with similar graphs from many other reactors) in the USAEC's annual presentation to the JCAE and, according to the rules, assigns & brief, descriptive reason for . each shutdown of 5 days or more. - A great deal more detail as to the nature . and cause of these and shorter interruptions in operation is given in the ) figures and 11 tables which follow. Figure L-2 covers 1966, the first year of po#er operation;- It shows vhen salt was in the fuel loop, when the reactor was at power, and assigns & number to each interruption in either. A number in a circle means the interruption was due to some experiment. A number in a square meané that the'interruption was necessary because of trouble with some system or com- ponent. Figures b-3, 4-h, and 4-5 display the same kind of information for 1967, 1968, and 1969 respectively. | In Tabies L-7 through 4-10, the interruptions in operation during each year are.categorized as to the type of interruption, the cause, and other work that was accomplished during the interruption. Each interrup- tion listed in these four tables is described more fully in Tables L4-11 through U-1k. (:W 23 Teble 4-6 Time Critical and Time Available During 233U Phase Of MSRE Program (September 10, 1968 - December 12, 1969) Hours Percentage Overall (9/10/68 - 12/12/69) 10959 | Periéd; | 7: N Elapsed Crit. Avail. Crit. Avail. Zero-power expts.’(9/10/68; 11/28/68) 1895 649 - 1852 3h.2 9T.7 Interim (11/28/68 - 1/12/69) 1092 8 110 7.9 10.1 First Power Runs (1/12/69 - 6/1/69) 3355 -~ 3175 3268 9.6 97.k Shutdown (6/1/69) - 8/9/69) 16k 0 8 0 24,8 " Later Power Runs (8/10/69 - 12/12/69) 297h ~ 2230 2971 T75.0 . 99.9 6140 8609 56.0 78.6 120 110 100 70 CUMULATIVE THERMAL MEGAWATT—HOURS (in thousands) 3 20 10 ORNL-DWG 73-691 AN = oo, h 10. 11. 12, 13. 14. 15. 18, 17. . Remove plugged capillary, check valve and filter . Repair electrical wiring short in fusl sample enricher, . Leak—test reactor cell. . Remove and replace core—sample array; replace ons . Clear ges line plug caused by salt overfill; replace . Clear salt plug from gas line; repair air valve in . Replace leaking sir-line disconnects in reactor cell, MAJOR SHUTDOWN PERIODS Low power testing. in offgas system. Identify plugging material and install larger valves and newly designed filter in offgas system, blower. second blower, reactor cell. MSRE - Initial Criticafity Critical on 233y Installed Capacity Reactor Thermal Mw Power Generation . 1989 Thermal Mwh TYOTAL TMwh 6/1/65 10/2/68 8 33,283 105,737 s/ Remove and replace core—sample array; annual tests of containment and controls. 131_4_/ Replace fuel sampler drive; retrieve sample fatch, Remove and replace core—sample array. Annual tests. Shake down 9533"""' plant. Process salt to remove U. Load U. Mix sait. Repair component cooling blower. Repair fuel sampler drive. Service control—rod drive. Remove and replace cors—sample array, Replace control rod and drive. Clear offgas line. Annual tests. : Measure fission product deposition. Conclude nuclear operations 12/12/69. 10 A I . ' t . ‘ ¢ ' i 4 i - : r7/'- _ /...fi..l K 2 3 ,/?/ 1965 Dec 1966 Dec 1987 Dec 1968 Dec 1960 Dec Fig. 4.1 1Integrated Power and Major Shutdowns of the MSRE C we 25 INVESTIGATE OFF-GAS PLUGGING — CHANGE FILTERS AND VALVES LOW-POWER DYNAMICS TESTS GO TO - FULL POWER REPAIR SAMPLER CHECK OPERATE - AT POWER CONTAINMENT et e e o s 1 REMOVE CORE SPECIMENS REPLACE MAIN BLOWER THAW FROZEN LINES TEST CONTAINMENT OPERATE AT POWER REMOVE SALT PLUG CHECK CONTAINMET OPERATE T POWER ORNL-DWG 73-592 “.POWER (MW) BRFUEL o ~ JFusw . o | o z3 | c pEEEl R , 20:0:0:0:0:0:0.0 2 15 31 15 ‘ | 1966 JAN FEB MAR APR | MAY JUN JUL AUG SEP ocCT Fig. 4.2 1Interruptions of Operations — Year 1966 ORNL—-DWG 73-593 HSN14 13Nnd (M) 43mod xQ 0 < ~N . n‘-- T iF NI LIVS 1. ™M . QO - o O ™ & o w0 3 xcHo - > WOT3 9 USIIG SUT welg ItV 4usmmI}sul sxayesl pue i Te0Ta399TE - = adeno xem0d VAL UOTHUTTIUSA % FUSTUTBIUCY mayehg JUBTOOD| _ : .unm_noaaoo : ) x umeds IaEy| - Cause and Related Activities wy 8hg - 883330 o - = sasmoTg 10 ‘ _10795p%Y : ‘ = sma18L5 3188 juBTo0) % Yend INTERRUPTICNS OF OPERATION DURING 1968 JOJIXH flflgm 38T ¥0ou) Table 4-9. *3dxg 03 snp anq ‘psuuerdun pomusta] A A A B Aoma B RoRe R uysaq 3ueTO0D| s upsaq Tend : , xEE XX ws1og peot| - ‘ = ‘ ol v o 8 wexos POy o w = % uopjonpay| sonoq| X E R T E *Report'a'bie' Unscheduled Rod Scram - = Automatic. - Primary | 2 | 898 -NES8A5LIE8S | 2| SRSSRSRSRRSIaaY No.. H M MmS INO DO NG Ay MmN W = Worked on During Period Following the Interruption; Includes Preventive Maintenance M = mnual 8 = Becondary A P = 32 Table k-10. INTERRUPTIONS OF OPERATION DURING 1969 What Bappened Couse and Related Activities g 2.0 g 5 Kl g 3 gf - 1 » - MARHAEHALH R RN T R sd) 20l 0 T 28 5 | g e dh ae] s 0] s dal a2 2 2 | | () o) ~e (BE) 3| 3] 3 8|4 8e) 3| 035 350880 5 HLEEER sl kel k| A B BB a 1 113 | u W 2 /23 | w P 3 1/26 M P A 1/28 | w P S 2/7 “ P 6 2/ M P. T 2/16 | M P 8 2/ M P-W 9 221 | 'm P ] 3n A oa P 5 1 3/8 N P v 12 3/20 x P 13 3/25 N P s | 3/2s A*] s P 8 15 k3 u P 16 | &/10 Ml a|afa 8 v 17 2 | x A - P 18 | a1 A"l A a 8 P-¥ 8 19 5/% N U 20 5/5 N W 21 5/13 ¥ P sf19 | x P 23 s/2s N P b 5/26 N P 5 5/26 A"l P 8 s/er ¥ A P 8 6/1 » I'n | x |x r 8V W v v v |Pvw 8/0 | ) 1 e 8 ' 29 8/z0 |. P 8 30 8/20 N P 8 . 31 8/20 X P s 2 8/20 N P 8 33 8/21 M P 8 3% 8/21 Y ? 8 35 | 6/28 A" P 8 6 | 8/29 A" P 8 BT | 8/29 | m P ’ B8 9T u P B9 { 9/9 X P o 9/16 N P b1 9/18 u -V h2 9/23 | m W h3 9/30 A P-W kb | 9/30 A P hs |~ 9/3% A W b | 9/3% A PV h7 9/30 A P-W b8 | lo/1 N P ho | 10/3 M P pO 10/15 M P 51 | 10/17 M P b2 | 10/23 | a C 8 P 53 | 10/2h A 8 P by | 11/2 ¥ | |xw x| b5 | 12/3 n , P b6 | 12712 |- w tn |nw {x|e v W *Reportable Unacheduled Rod Scram A = Automatic M = Manual . P = Primary 8 a Becondary Y ¥V = Worked on Durinz Pericd Following the Interruntion; Includes Preventive Naintenance Teble 4.11 33 Description of Interruptions of Operation " During 1966 | | o No. - Date " Description and Related Activities 1° 1/5-1/20 2 1/21 3 1/22 Y 1/23 5 1/23 6 1/23 | A T 1/24 8 1/25 9 1/25-1/26 10. 2/16 -;Low—power tests to callbrate nuclear instruments, check radiation levels, collect data for &ynamlcsl‘ and noise analysis, obtain reactivity balances,-- and check out the control circuits. Rod scram due to ac01dental shortlng of the 32-¥ dc' power supply while 1nvest1gat1ng a No. 3 safety channel trlp Rod and load scram to test ecircuitry. Rod scram due to a voltage sag. - . Pcwer-reduétion to attempt to blow out plugs in off- gas line, equalizer 1ines,'and.main-charcoalfbeds. Rod scram.due to 1nstrument malfunctlon when SW1tch1ng linear channel ranges. Load scram due to false low coolant flow indication. Rod scram due to inadvertent actuation of the manual scraméswitch Fuel and coolant drain due- to plugging in the offgas: system and to insulate the coolent flow transmitter 1ines.‘ While shut down, the restrictor in the equa-. lizer line (521) and the check valve in the vent line (533) to the auxiliary charcosal bed were re- moved. The filter in the main offgas line was re- placed and-the pressure control valve (PCV-522) was replaced with a hand velve with larger ports. Work was done.on No. 3 control rod drive, component coolant pump No. 1, radiator door seals, and motor generator No. 4. The thermal shleld air lock = prdb1em.was investigated. Fuel drain due to a fallure of the draln tank cell ‘space cooler motor, . While shut down, the inlet. ~ valves to the charcoal beds were ‘replaced and the offges line at the fuel pump was reamed out. Re- pairs were made on a damaged electrical penetration on the’ sampler-enrlcher and a8 1eak on reactor cell % - , Reporteble unscheduled rod scram. 34 - 20 Teble 4.11 (continued) . | - B | kifi No. Date Description and Related Activities 10 (con't) space cooler No. 2_wds‘fixed. A static inverter was installed to replace motor generator No. 4. The treated water corrosion inhibitor was changed from _ potassium to lithium nitrite-tetrsborate. 11 L/6 Fuel drain due to a plug at the auxiliary charcoal bed ‘ which turned out to be a poppet from a drain tank vent check valve. 12 /12 Rod scram andiload_SCram to test performance of the rods ' and radiator doors. 13 L/14 Power reduction to collect data for noise analysis. * ) 14 4/16 Rod scram due to the fuel pump stopping which was caused by a false low salt level 1nd1cat10n. 15 L/20 Manual rod scram and. load scram to end the 2 S=Mwr operation. ¥ _ _ | o 16 L/20 Rod scram due to a false signal on safety channel No. 1 - while safety channel No. 2 was being tested. ‘i; 17 Lh/21 Power reduction due to charcoal beds plugging. * ' | o | , 18 y/22 Rod scram, load scram, fuel drain and coolant drain initiated by a faulty pump pressure relay which tripped instrument power breskers. - Some salt froze in radiator because the cooclant pump could not dbe started due to a lubricating oil flow switch | failure (FS-T54). . _ - , 19 h/25 Rod scram due to inverter failure. . , _ L/25 Rod scram and load scram due to false low fuel-pump level indication whlle doing periodic instrument | check llsts. . | 21 4/28 Rod scram and load scram due to TVA power outage | caused by & storm. # ' _ . 22 4/28 Rod scram and load scram due to spurious safety chan- nel trips possibly caused by an electrical storm. 23 h/29_ ‘Manual rod scram and load scram to check effect of cold doors on radiator temperatures. Shutdown was necessary to repair the latch on the sampler- o enricher. - b’ ' ® Reporteble unscheduled rod scram. 35 Table 4.11 (continued) 132 3 No. Date Description and Related Activities 2k L/29 Fuel drain necessitated by a failure of the samfiler- enricher cable drive motor. While shut down, the belts were tightened on component cooling pump No. 1 and attempts were made to blow out the plugs in the offgas lines. 25 5/11 Power reduction due to offgas line filter plugging. % ' > 26 5/12 Rod scram and load scram due to operator failing to in- sert a control circuit jumper while doing perlodlcal instrument check llsts. 27 5/16 Power reduetlon to increase pitch on main blower blades to increase possible maximum power. # . 28 5/19 Rod seram and load scram due to a TVA power outage. While shut down, measured offgas pressure drop and attempted to blow out plugs ¥ 29 5/23 Rod scram caused by instrument malfunctlon end operator error. Operator had switched to servo with a LO°F difference between actual temperature and temperature demand setp01nt. ' 30 - 5/25 Power reduction to check the power coefficient of Te— act1v1ty. o : 31 5/26 Power reduction to obtain reactivity balance data. 5/26 Power reductlon due to main blower No. 1 stopplng and falllng to start. 33 . 5/26 | Power"reduction to collect data'for_anoise analysis. F * . el T . . 34 s5/26 Manual rod screm &nd load scram to investigate a R false flre alarm. 5/28 Fuel draln due to & failure of component c¢oolant pump - No. 1 while component coolant pump No. 2 wes valved off. - A planned drain had been scheduled due to the ~ high indicated cell leek rate. This was found to be due to,arleak from the thermocouple headers into the cell. :While shut down, a 20-psig leak test was made ¥_ — RCITE , Reporteble unscheduled rod scram. 36 Table 4.11 (continued) No. Date Description'and Related Activities 35 (con't) on the cells, repairs were made on both 6omponen£ coolant pumps, the electrical load was redistributed and attempts were made to locate the water leak into the cells. 36 6/1h Power reduction due to & failure in the electrical supply to the nuclear instruments. 37 6/1k Power reduction due to failure of a flexible coupling : on main blower No. 1. 38 6/19 Power reduction to empty the overflow tank. € - - : 39 6/20 Rod scram and load scram due to the operator inadver- tently inserting a test source at the wrong process radiation detector (RE-528 instead of RE-565) while ‘ doing periodic instrument check lists. * - ' ‘ 40 6/27 Rod scram, load scram, fuel drain, and coolant drain due to the electrical supply to component coolant pump No. 1 shorting out at the penetration. . | | , L1 T/14 Rod scram and load scram due to a TVA power outage. ¥ 42 T/15 Rod scram and load scram due to a TVA power outage. caused by a storm. 43 T/17 Power reduction and coolant drain due to & catastrophic failure of main blower No, 1 hub. The fuel salt was drained on T/24/66. While filling the fuel system - - with flush salt, the fuel-pump bowl was accidentally overfilled and some salt froze in the sampler and offgas lines. Heat had to be gpplied remotely to remove the plugs. While shut down, the core speci- mens were replaced, and a new particle trap was in- stalled in the offgas line. Lesks in reactor cell space cooler No. 1 and in the treated water heat ex- ‘changer were repaired and the thermal shield degas~- sing tank was installed. Both component coolant pumps were repaired and piping was installed to drain the condensed wsater from them. Repairs were made on the radiator door seals, radiator heaters, and the main blower motors. All the electrical switch gear breakers were calibrated. Prior to startup, the fuel and coolant pumps lube oil was changed, the periodic check lists were completed and the cells were leak-tested at 10 psig. ¥ Reportable unscheduled rod scranm, 37 Table 4.11 (continued) No. Date Description and Related Activities bl 45 # Lo h7¥ 48 ko 50 5 22 53 54 % 55 10/10 10/10 10/12 10/16 10/31 - 10/31 11/11 11/12 11/15 11/17 11/20 12/23. .- Power reduction due. to a false signal from a process radiation detector (RE-528). Power reduction due to an experiment when the fuel-pump pressure was rapidly vented from 15 to 3 psig. Rod scram due to inverter trouble. ‘Rod scram and load scram due tb'false signal from a process radiation detector (RE-528). Power reduction to install & new rotor on main "blower No. 1. 'Fuel drain due to upper offgas line (524) being plugged. While shut down, the flow restrictor in this line was replaced and heat was applied to main . . offgas line at the fuel pump to unplug it. Power reduction to check main blower vibration. Power‘fEduction‘to empty the overflow tank. Power reduction to empty the overflow tank. Power reduction to attempt to blow the plug out of the main offgas line at the fuel pump. ) Fuel drain due to a high indicated cell lesk rate which proved to be due to lesking air line disconhects on .. in-cell valve operators. Rotemeters were installed on these air lines. While shut down, the plug in the offgas line at the fuel pump was reamed out and ' repasirs were made on the component coolant pumps and “the component coolant system pressure control valve (PACV-960). Hesters were instelled on the inlets of - main charcoal beds 1A and B. The cells were leak- tested at 10 psig. = - Rod scram and load scram by accidentally opening a cir- - cuit while installing jumpers for an experiment. % o Reporteble unscheduled rod scram. 38 Teble 4.11 (continued) No. Date Description and Related Activities 56 12/2h - Power reduction caused by a pressure release experi- - : - ment. The reactivity decreased, the regulating rod withdrew to its upper limit and the operator failed to change the limit. 57 12/26 Power reduction when venting the fuel pump after an attempt to blow out the plug in the mein offgas line. The regulating rod reached its upper limit and the operator failed to change the limit. & , . Reportable unscheduled rod scram. 39 Table h 12 Descrlptlon of Interruptions of Operatlon Lo Durlng 1967 No. Date Description and Belated.Acfi#ities_ 10 1/12 1/1k 1/16 1/28 1/30f2/2j 2/5 o 2/26-2/2T 3 ii'3/6 3/7 Rod scram end load scram due to a freeze valve (FvV-108) relay tripping the reactor out of RUN during an in- vestigation of the plastic keepers on the relays. Power reduction to cbserve xenon decey rate and ob- tain zero-power reactivity balance data. - " Fuel drain due to leskage from the in-cell air line disconnects. The measured leakage was so high that _ reasongble cell lesk rate calculations were not pos- ‘sible. The disconnects were replaced. While shut "down, the main offgas fllter and: control valve (PCV-522) were removed and a new filter was installed. ~ The treated water heat exchanger was replaced and . maintenance was done on the component coolant pumps. ~ The cells were,leakftested at 10 psig. Power”feducbiofi!to.feet automatic load control. ','Power reductlon to take plctures of the radiator and 1nvestigate heat transfer of the radiator and heat exchanger. “Also made dynamic tests and checked the temperature servo system. : : Power'reductlon due to the'failure of a freeze valve module (FV-107) which trlpped the reactor out of RUN mod.e. B : Power reduction due to main blower No. 1 vlbratlon. cPower reductlon to-dbtaln,heat—transfer data.; Power reduction due to coolant system offgas filter plugglns - “l Pcwer reductlon due to electrlcal power supply trouble to nuclear safety and rod servo. . 'lPower reductlon due to 8 bad bearing on main blower - No. 3. Remained at low pover to observe effect of xenon decay on resctivity balance * ( Reportable unscheduled rod scram. Tsble 4.12 (con't) 40 No. Date Descriptibh and Related Activities * 12 13 1k 15 16 17 18 19 20 21 22 3/13 3/19 3/23 411 “h/28-4/30 5/5 5/8 6/21 6/23 6/25 - 6/30 Rod scram and load scram due to false signal on safety channel 3 while doing perlodlc 1nstrument check lists. Rod reverse and load scram due to false low fuel-pump speed indication while doing periodic instrument check lists. ' Power reduction due to vibration on main blower No. 3. Power reduction to obtain a special fuel-pump gas sample. Also obtained heat-transfer data. Power reduction to cdllect date for noise analysis. Power reduction to test reactor response to load | changes W1thout rod changes. Rod scram, load scram, fuel drain and coolant drain to remove reactor core specimens. While shut down, . gamma-scan data were taken, a new source was in- stalled, and heaters were installed on the inlets to main charcoal beds 2A and 2B. Reactor cell space cooler No. 2 was replaced and repairs were made on control rod No. 2, radiator door brakes, component coolant pumps and the main blowers. A boot was re- placed on the sampler-enricher. Prior to startup, the periodic check lists were completed, the cells wvere leak-tested at 20 psig. Power reduction to'obtain heat-transfer data. Power reduction to repair an oil pressure switch on component coolant pump No. 1. A Rod scram and load scram due to a TVA power outage caused by a storm. While shut down, repairs were made on component coolant pump No. 1 oil pressure switch, a treated water rupture disc (84h) was re- placed, and a heavier base was installed on main . blower No. 3. ‘Rod scram and load scram due to an operator resetting the wrong channel while doing periodic instrument check lists. ¥ | Reporteble unscheduled rod scram. Table L4.12 (con't) 41 No. Date Description and Related Activities o 23 T/12 Rod scram and load scram due to a TVA pover outage ) | caused. by a storm. oh 8/T Power reduction to obtain heat-transfer data. 25 8/8-9/11 Power reduction, fuel drain and coolant drain dué to the sampler-enricher drive cable being severed. During shutdown, the latch was retrieved but the capsule remained in the fuel pump. Prior to startup, | ' maintenance was done on compohent coolant_pump No. 1. 26 9/11 Coolant drain due to the radlator door blndlng in the ' guides. 27 9/18 - Power reductlon due to an 011 lesgk in component coolant pump No. 2. | 28 9/18 Fuel drain due to a leak through the discharge valve of component coolent pump No. 2 which prevented repair of the oil leak. 29 10/1 Power reduction to collect date for noise analysis. % o ’ - 30 10/20 Rod scram and load scram due to a power failure due to ‘an arc between a switch actuator rod and the main power llne to the building. 31 :10/23 Power reduction to observe the effect of xenon decay : on the reactivity balance. . 32 11/1k Power reduction to collect data for noise analysis. 33 kéll/l6 'Pofier reductlon to repalr sampler—enrlcher wiring. 3 11/22- Rod scram.whlle malntenance was being performed on the B rod servo instrumentatlnn.- - 35 ~f 12[13 Power reductlon to Observe the - efféct of xenon'p01son— ing on the reactivity balance and to collect data for noise analy31s._ # L Lo . Reportable unscheduled rod scram. 42 Table 4.13 Description of Interruptions of Operation During 1968 Description and Related Activities No. Date 1l 1/22 2 2/21 3 2/29 Y 3/7 5 3/22 6 3/25 T 3/25 8 3/26 9 9/1k Power reduction to observe the effect of xenon poison- ing on the reactivity balance and to collect data for noise- analysis. Power reduction to test the rod servo instrumentation "under simulated 233y conditions. Power reduction to observe the effect of xenon poison- ing on the reactivity balance and to collect data for noise ansalysis. Power reduction to observe the effect of xenon poison- ing on the reactivity balance, to make dynamics tests and take heat-transfer data. Power reduction to meke dynamics tests. Manual rod scram end load scram to observe xenon poisoning effect on reactivity balance.: Manusal rod scram to end_zasU operation.' Fuel drain to end 23°U operation. Shutdown was emmi- nent due to & tangled drive cable in the sampler- enricher. After the drain, the capsule was acci- dentally dropped into the fuel pump when trying to untangle the cable. Unsuccessful attempts were made to recover this. While shut down, gammas-scan data was teken, the core specimens were removed, and the main offgas line . gt the fuel pump was unplugged. - The flush and fuel salts were transferred to the ‘£Uel processing plant where the 235U was removed. 33 was added to the fuel salt. Repairs were made on two heat exchanger heaters, control rod No. 2, both component coolant pumps, the inverter, and the radiator doors. Meain blowers No. 1 and No. 3 bear- ings were replaced and the coolant offges filter wes replaced. Prior to startup, the periodic check lists were completed and the cells were leek~tested" at 20 psig. Fuel drain for adding more 233U as part of the criti- cality experiment. Table 4.13 (con't) 43 Déécription and Related Activities No. Date 10 9/17 11 9/21 12 10/21 , ¥ ' 13 11/27 1k 11/28 15 12/17 Fuel drain for adding more 233y as part -0f - the criti- cality experiment. : Fuel drain for adding more 233y as part of the criti- cality experlment. Fuel drain due to inverter fallure. Operatof’falled to reset reactor cell radietion monitor in time to pre- vent drain. Rod screm and load scram due to operator failing to in- sert a Jumper around fuel pump level contacts. The fuel-pump pressure was being vented after attempt- ing to blow out a plug in the main offgas line. Fuel drain and coolant drain to mix the fuel salt in the loop and the drain tank. The offgas line at the fuel pump was reamed out, the coolant offgas piping was cleaned, and repairs were made on the component coolant pumps. Fuel drain due to difficulty with the latch and cable drive of the sampler-enricher. While shut down, control rod drive No. 3 was inspected and a weight added to decrease its drop time. . | | R Reportable unscheduled rod scram, - 44 Teble L.14 Description of Interruptions of Operations | During 1969 ' No. Date Deseription and Related Activities 1 1/15 Power reduction to repair lower limit switches on radiator doors. 2 1/23 Power reduction to investigate'effect of power on "blips". _3 -1[26 Power reduction to infiestigate effect of stopping the fuel pump on "blips". k1728 Power reduction to investigate "blips". 5 /7 Power reduction to collect data for noise analysis. 6 2/11 Power reduction to run tests on natursal ponvection; T 2/16 Power reduction to run dynamics tests. 8 2/23 Power reduction due to failure of belts on stack f&n NO. lo ’ . 9 2/27 Power reduction to connect the fuel pump to a variable , speed motor generator set. * ! - 10 3/4 Manual rod scram and load scram due to failure of the veriable speed motor generator set. 11 3/8 Power reduction to -allow xenon to decay in preparation to making void fraction tests. While shut down, the coolant offgas filter was replaced. ' 12 3/20 Power reduction to connect the fuel pump to the normal electrical supply. 13 3/25 Power reduction to connect the fuel pump to the varisble frequency motor generator set. . 1 3/25 Manual rod scram and load scram due to failure of the variable speed motor generator set. 15 L4/3 | Power reduction to connect the fuel pump to the normal electrical supply. # _ - Reporteble unscheduled rod scram. Teble L4-14 (continued) 45 6/1 - No. Date Description and Related Activities. ¥ ’ . ’ . ’ L ) 16 L/10 Rod scram, load scram, fuel drain, and coolant drain due to a burned-out fuse on one phase of the main ~transformer. Drains could have been averted by operating alternate equipment. 17 L/12 Load scram due to a disturbance in the 48-v dec power supply. = ' 18 /15 Rod scram, load scram, and fuel drain due to the ' drain freeze valve thawing. The setpoint on a tem- perature switch had drifted off the setpoint and an | inadequate plug had been established. , 19 5/4 Power reductlon due to a failure of the coolant stack berylllum monitor. 20 5/5 Power reduction due to an error in calculating the reactiv1ty balance : 21 5/13 Power reductlon to connect the fuel pump to the variaeble speed motor generator set. 22 5/19 Power reduction to connect the fuel pump to the normal ' electrical supply. 23 5/25 .Power reduction due to plug in overflow tank vent line vhlch caused dlfflculty in emptylng the overflcw tank. ' 2k 5/26 Power reduction to connect the fuel pump to. the varlable speed motor generator set. o 25 - 5/26 Manual rod.scram and load scram due to. fallure of the : varlable speed motor generator set. 26 5/2T Manual rod scram snd load scram due to fallure of the o variable speed motor generator set. 27 - . Manual rod scram, load scram, fuel drain, and coolant ~ drain to remove core spec1mens and due to plugging in the offgas system. Control Rod No. 3 did not scram. It was replaced during the shutdown. Ex- : ten51ve gamma-scan date was tasken. A heater was in- stelled on the mein offgas line near the fuel pump which aided in removing the plug. The overflow tank * Reportable unscheduled rod scram. 46 Table L4.14 (con't) No. - Date | Description and Related Activities: 2T {con't) offgas line was replaced and the coolant offgas sys- o tem was unplugged. Repairs were made on the sampler- enricher, and on a leaky in-cell air line disconnect. Before startup, the periodic check lists (modified) were completed and the cells were leak-tested at 20 psig. Full power was delayed while tests were made at various fuel-pump speeds to 1nvest1gate bubbles 1n the loop. 28 8/20_. Power reduction due to failure of the variable speed motor genersator set. 29 8/20 Power reduction due to failure of the varisble speed motor generator set. 30 8/20 Power reduction due to failure of the variable speed motor generator set. 31 8/20 Power reduction due to failure of the variable speed motor generator set. 32 8/20 Pover reduction due to feilure of the variable speed motor generator set. - 33 8/21 Power reduction due to failure of the variable speed motor generator set. 34 8/21 Power reduction due to failiure of the variable speed : motor generator set. 35 8/28 Manual rod scram and load scram due to failure of the variable speed motor generator set. > * . _ : ' 36 8/29 - Manual rod scram and load scram due to failure of the variable speed motor generator set. 37 8/29 Power reduction for xénon decay in preparatibn for argon cover-gas experiments. 38 9/1T Power reduction due to an error in calculating the reactivity balance. 39 9/9 Power reduction to collect data for noise analysis, to meke gamma scans and run rod drop tests. Lo 9/16 Power reduction to connect the fuel pump to the normal electrical supply. * . Reportable unscheduled rod scram. l}r\ Table 4.1k (con't) 47 No. Date Description and Related Activities 41 9/18 Power reductlon to replace 8 bearlng on Main Blower i No. 1. b2 -~ 9/23 Power reduction to replece a bearing on Main Blower No. 1. While shut down, connected the fuel pump to the variable speed motor generator set. 43 9/30 Load scram due to dirty contacts on relays. L)y 9/30' | Load scram due to dirty contacts on relsays. L5 9/30 'fLoad scram due to dirty contacts on relays. h6 9/30 Load scram due to dirty contacts on relays. L7 9/30 Loed scram due to dirty contacts on relays. L8 10/1 Power reduction’fo connect tfie fuel pump to the normal - electrical supply. b - .h9 10/3 Power ?eduction due to door being insdvertently left ' ’ " open in an exclusion area which caused inadequate containment ventilation. - 50 . 10/15 Power redfiction to take 8 special fuel-pump gas sample. 51 10/17 Power.reductien to take special ffiel—pump offgas sam- : ples and to run tests with the fuel pump off. 52 10/23 Load scram due to a false trip of a coolant salt flow ' B reley while doing periodic instrument check lists. 53 10/2k | Laad'Seram'while7testing'fe1ays after they were re- paired. 'Operator failed to reset'the relays. 5k 1/2. Manual rod ‘scram, load scram, fuel draln and coolant : - draln to ‘teke gamma-scan data, 55 12/3 Power reductlon to take gamme-gcan data. 56 12/12 ' Manual rod scram, load. scram, fuel drain, and coolant drain. tQ end operation of the reactor. 48 The next three tables summarize the foregoing mass of 1nformatlon on interruptions, Table 4-15 is a summary by type; Tsble h—16 by cause and work done during the interruption. Because unscheduled scrams of the con- trol rods are of special interest, a separate breskdown of these is given in Teble 4-17. It should be noted that never were the rods seremmed because 7 of the reactor power, the #eactqr peribd, or the fuel temperature going out of limits. 4,4 Time Required for Operstional Tasks , One aSbect of the opefdbility of a plant (and ohe which affects plant . availdbilify) is how long it takes to do various tasks required in the opere- tion. The purpose of this section is to summarize the MSRE experience in this regard. ' Manpower and the approkimate number of working hours reqpired.for verious tasks are listed in Teble 4-18. One must recognize that some of these figures are subjeet to considerable vaE}ation. The time required to do a particular task'depended on chance difficulties that were encountered, the experience of the crew, other jJobs being done concurrently and the urgency of speedily completlng the task, i.e., whether or not it was given a hlgh frlorlty. Furthermore, sometimes extra requirements such as special data-teking caused the Job to take longer than usual. The figures in Tsable 4-18 apply to fairly normel situstions; not the best nor the worst. Very brief descriptions of the tasks.are given below, with references to the ‘section of the MSRE Operating Procedures2? followed in each ease,rif applicable, L.4.1 Auxiliary Systems Startup Check Lists , During a shutdown, much of the equipment remained in operstion. However, after long shutdowns, the auxiliary systems startup check lists (Section 4) were completed to assure that nothing had been overlooked. All valves and switches were checked, eqfiipment was put into operation and standby equipment' was tested. These were done as late in the shutdown as practical. LA — Electrical Startup Check List assured that all main breskers were closed so that equlpment and heaters could be started from the control boards. ' 49 Teble 4-15 = Summary of Interruptions by Type Number of Interruptions Type of_Iptérifiptiohf:_ - . 1966 1967 1968 1969 Total Power Reduction: Automatic 3 -0 0 0 -3 | Manual 23 23 D 39 90 Rod Scram: Automatic 20 T 1 T 35 | : Manual . 6 1 2 3 12 Load Scram: ~ Automatic 13 8 1 1h 36 | Manual 6 1 1 3 11 Fuel Drain Automatic 3 1 2 6 | Manual T L 6 3 20 Coolant Drain Automatic 1 0 0 1 2 - Manual 3 3 1 3 10 Total - 85 Yy o 18 75 225 50 . | Table 4-16. Summary of Interruptions by Cause and Work Done During Interruption - Number of Interruptions - 1066 - | __1067 ]_1968 |__ 1969 . Total Category P S W|P s wlp S WI|P S WI|P 8 W Planned 8 o o1k 0 0fl2 o 02k 0 0|58 O O. Unplanned | But Due to an . : Experiment 3 00 O 0 0OjJO0O O0.0J13 0 0116 0 © Check Lists o 50| 120|000}010}|1 8 ,_ Human Error T 1 0 21 011 0fj2 3 011 6 3 Coolant Salt | o | . Systems '3 00| 000]0O0OO0O]|]0OO11f3 01 Radiator or , _ fl Blowers 6 1 5 L o 5|0 0 12 0 3|12 113 Offgas Bsytem {10 o012 0 0 0 2 1 2 |12 118 Water System 1 o 4! 01 3[000 ofl 1 7 Component Coolant ? | § System 2 0 6 2 0 210 0 210 0 O{ 4 010 Containment eand =~ -~ = - i ' | _ Ventilation 1l 1 2 O 0 0/0 0 O 1 0 12 1 3 |TVA Power Outage 3 1 0 2 0 0|0 0 0 0 0 0|5 1 O Electrical and | . Heaters 4 o-4:; 1 0 1{0 01 ;213 1713 T Instrument Air ! § ' g System 1 01{ 101|000;001[2 03 Instrumentation 11 o0 3| 70 7|1 o0 o01{11 0 930 019 Control Rods O 02, 002/002;001|00T Freeze Valves : O 00: 00 0jO O O}O0 10O 10 Samplers 2 03 2. 03{112}001(5 19 Periodic Con- * - | tainment Tests 0 03] 002]0oo011i00 112|007 |Core specimens | 0 o0 1| 1 0 ©01j10 2 0 4 NOTE: P = Primary cause of interruption. S - = Secondary cause of interruption. W = Worked on during period following the interruption (includes preventive maintenance). % _ No interruption of operations during this entire period was caused by the cover-gas system, lube-oll system, or leak-detector system. Table 4.17 MBRE Cumulative Cycle History _ Thaw Usage ' . ‘ ' Quench & Factor " Component. Heat/Cool|Fill/Drain|Pover |Quench | Time | On/Off| Thaw| Trans.: & *These figures are based on the original calculations, If they were based on freeze flange thermal cycle tests, the usage factors would be 23.04% and 13.37%. - TI6 Table L4-18, APPROXIMATE TIMES REQUIRED TO PERFORM VARIOUS OPERATIONS ‘ _MANPOWER Operating ' E= Engineer Procedures Time to T= Technician Section Complete P= Pipefitter Number Operation (hrs) I= Instrument Mechanic LA Electrical startup check 1list 10 - 15 1-T*‘- 4B Instrument air system startup check list L -8 1-7" ke Water system startup check list 4y -8 -7 4D Component cooling system startup check list 1 -2 l-T* Ll ~ Shield and Containment startup check lists . N o (1) Ieak test of containment valves ‘130 - 160 1-E, 1-7, 1-2* (2 Pressurize the cells (to 20 psig) 15 - 18 e (3 .Cell leak rate at pressure >60 - e (4 Evacuate cells and purge with nitrogen 18 - 24 e (5) First cell leak rate at -2 psig S170 W LF Ventilation system startup check list L - 8 1-1 hg Leak detector system startup check list 2 -4 1-7" LH ‘Instrumentation startup check list. 160 - 200 1-E, 1-T, 1-I SA Purging oxygen and moisture from the fuel circulation system 30 - ko L 5B Startup of cover-gas and offgas systems - ‘ 2 -k l-T* . * 5E Startup of lube oll systems 2-4 1-T 5F Heatup of fuel and coolant systems 4o - 50 SR S5H Routine pressure test 20 - 30 NN * . Done on shift, *fMost of this was done on day shift with some assistance from the shifts, ***This d1d not require much attention. This required the attention of most of the shift operating crev, cs Table L4-18, APPROXIMATE TIMES REQUIRED TO PERFORM VARIOUS OPERATIONS (continued) - - MANPOWER Operating o o : : E= Engineer Procedures . | Time to T= Technician Section ‘ ' Complete P= Pipefitter ~ Number ‘ Operation ' (hrs) I= Instrument Mechanic 51 Filling the fuel and coolant systemé © (1) Filling the coolant system ' - 9-12 XK (2 Filling the fuel system with flush salt 9-12 FHHX (3) Drain of flush salt and secure - 3-5 *hRx () Prepare for fuel f£111 - . 8-9 RN - (5) 'Filling the fuel system with fuel salt | | 10 - 12 RN 5 | Criticality and pOWer operation o | ‘f(l)V | Preparation for. ‘eriticality and power operation : 1-2 Ty (2) Suberitical to full power. , | 1 1-7 6A3 or I | o j . . ‘ ‘ 6Ah Fuel salt sampling or enriching 3-4 2-7 6B Coolant salt sampling 1 1-T 8 . Periodic instrument checks 8A '~ Neutron level instruments 5 -7 2-1 8B Process -radiation monitors _ 1 3-7 - 8D Safety eircuit checks 2 =L 2 to 3-T o Emptying the overflow tank 1/2 -1 e 124 Logs | - 12A-2A First control room log on each shift 3/4 - 1-1/4 'l-T* * Done on shift. Most of this was done on day shift with some assistance from the shifts, This did not require much attention, : This required the attention of most of the shift operating crew, ¥ i £q Table 4-18, APPROXIMATE TIMES REQUIRED TO PERFORM VARIOUS OPERATIONS (continued) MANPOWER Operating . B= Engineer Procedures Time to T= Technician Section ' Complete P= Pipefitter Number Operation (hrs) I= Instrument Mechanic 12A Logs (continued) ' 12A-2A Second control room log on each shift 1/6 l-T: 12A-2B First bullding log each day 1-1/2 - 2 1-T, 12A-2B First bullding log on other shifts | 1/2 - 1 1-T, 12A-2B Second building log on each shift 1/3 - 1/2 1-T 10 'Shutdown of the Reactor | (1 Full pover to subcritical 0-1 1-T, 2 Fuel salt drain and secure _ 3~-5 1-T* 3 Coolant salt drain and secure 2-14 1-7 L - Cool down to LOO°F L8 - 60 6k OA-1 Restarting equipment after an electrical power outage 1/6 N , * 5J-1 Return to power after a load and rod scram : 1 1-T 114, sI, | | & 5J Return to powver after a drain (1) Drain to both drain tanks (transfer through transfer lines) 50 - 60 WHHH (2 Drain to both drain tanks (transfer through £ill lines) 20 - 30 R (3 Drain to one drain tank _ _ _ 10 - 15 RN . Done on shift. , **Host of this was done on day shift with some assistance from the shifts. **¥IMis d1d not require much attention, O s required the attention of the shift operating crew, 1S 35 4B — Instrument Air Systeni Startup Check List assured that the com- pressors and driers were opereting properly, that the standby compressor would automaticallj start if needed, and that emergency nitrogen cylinders were installed. o . LUC — Water System Startup Check List assured that all pumps and tower fans were operable and that proper water flow was established on all equipment. 7 | - ‘ ) 4D — Component Cooling Systems StartupdCheck List assured that all vaiving was set properly and that all blowers were operable. UYE — Shield snd Containment Startup Check List is divided into 5 parts. Part (1) checked that the primary and secondary block and check valves, containment enclosures, and llnes were leak-tight to assure ade- quate containment. Removable' shieldlng was also checked to assure that it ‘ had been reinstalled. This check list was assigned to an operatlng engi- neer on day shift with technic1an and plpefltter help as needed. Part (2) — this is the time required to pressurlze the cells to 20 psig by adding instrument air. Part (3) involved taking pressure and tempera- ture data periodically to establish the cell leak rate. Some containment valves and lines were checked while atrrressure. The time required for checking these is included in Part (1). Part (L) is the time required to vent the cells, evacuate tcf—2.psig, and purge with nitrogen. Part (5) alsc,involved taking periodic data. Since there was a small wster leak in the cell,'it took about T days for the atmosphere to reach equilibrium hF —-Ventilation System Startup Check List assured that a stack fan was in operation, that the other one would automatlcally start and that ventllatlon was adequate in all aress. o hG —-Leak Detector. Systems Startup Check List assured that all valves were open to leak-detected flanges.and that the overall leak-rate was satisfactory. e | i ' hH —-Instrumentatlon Startup Check List assured that each instrument 'and circuit functioned properly Abnormal conditions.were similated and where possible the control acticn'was'ellowed to occur. One operating engineer on day shift wasraSSigned respohsibility-for completing this - check list with technicien and instrument mechanic assistance as required. Some of the tests were done on shift. 56 4.,4.2 Reactor Startup The manipulations involved in actual startup of the reactor (Section 5) are described below. ‘ | | SA — Purging Oxygen and Moisture from the Fuel Circulating System in- volved setting the helium purge at maximum rate and opera&ing the fuel pump to circulate it. The time of purge was a calculatedvnumber. 5B — Startup of Cover Gas and Offgas System involved setting the valvgs'and checking the helium treatipg station. o _ ’5E_—-Start of Lube.dil Sysfems involved checking the supply tank, draining the o0il catch tank, éetting valfes and checking that the standby pumps would start sutomatically. | | ' SF — Heatup of Fuel and Coolent Systems involved raising the heater settingé end maintaining a reésonéble temperature distribution. The fuel - and coolant systems could be heated simfilfaneously. The time given is for heating the coolant system from room temperature and the-reactor sys- tem from 400°F (normal shutdown condition) to 1200°F ahd reaching equilibrium. 5H — Routine Pressure Test involved raising the fuel system pressure to 60 psig (usually with flush salt circulating), then lowering the fuel system pressure an@ increésing'the coolant system pressure to 60 psig. All pressure switches and their control or alarm actions were checked. 5I — Filling the Fuel and Coolant Systems is divided into 4 parts. Part (1) involved thawing the freeze valves, filling the coolant loop, .checking thaw times of the freeze valves, refilling and starting the coolant pump. Part (2) involved thawing the freeze valve, filling the fuel loop with flush salt and starting the fuel pump. - Part (3) involved -dreining the flush salt, emptying the overflow tank and freezing the freeze valve. Part'(4) involved thawing the freeze valve, adjusting tem- peratures, running rod drop times and various other safety checks and ~taking base line countrate data. Part (5) involved filling the fuel loop with fuel salt, stopping at six levels to teke count rate deta, freezing the drain freeze valve and starting the fuel pump. - 53 —-Criticality‘and Power Operation is divided into 2 parts. Part (1) involved running rod drop and other tests. The time does not - maintaining containment. qu_samples could be delivered to the laboratory 57 include check lists 84, 8B, or 8D which sometimes had to be done before criticality (see 4.4.3). Part (2) involved menipulating the rqu‘and heat-removal system componentsrto reach full power. L. 4,3 Operation Various Jobs were done while the reactor was operating. These are described below. _ 6A3 or 6Alk — Fuel System Sampling or Addition of Enriching Capsules involved a series of;manipulations to insert and withdraw a capsule while " per shift if no difficulties were encountered. This rate could not be sustained due to decontamination bf carriers, etc. | | 6B — Coolant Salt Sampling involved a less complicated series of - manipulations to insert and withdraw a sample from the coolant pump. 8 — Periodic Instrument_Checks involved testing as much of the eritical equipment as possible without interrupting operation of the reactor. | ' : | 91 — Emptying the Overflcw Tank involved pressurizing it with helium and venting when it was nearly empty. This took about twice as long when the FP offgas line was plugged( 12A — Control Room and'Building Logs were taken tfiiéé per shift. Some adjustments were usually required, water treatment was added and other odd Jobs were done as part of the logs. k.%.4 Shutdown of the Reactor This (Sectlon 10) is descrlbed below. - It is divided into L parts. Part (1) involved reduc1ng the power. This could be done by scramming | the rods and the load or by 1nsert1ng the rods and lowering'the radiator doors. Part (2) involved-draining the fuel system and freezing the freeze "f,valves. Part (3)'ifivblved drsining the coolant system and'ffeezing the freeze vaIVes; Part (L) 1nvolved coollng the systems to hOO°F " '4,4.5 Recovery from Unplanned Shutdowns A numberrof times durlng operatlon, power outages and other unex- pected events caused shutdowns. Often there was a desire to return to operation as soon as possiblé. Some of these are described below. ORNL-DWG 73596 DAYS DAYS 1-30,~26 -20,~15 -10,-5 | 1 | 2 , 3 , 4 .8 , 8 , 7 4 B ; 9 , 0 ; M, 12 , B , 4 4 15 | i 1 | [ I | o i } I I I i Repairs, modifications, preventative maintenance and core specimen replacement. Instrumentation startup check list. E Containment startup check list, Other systems startup check lists. Last closure of primary system, Purge of primary system, Wald membrane and instal] blocks on cells. Pressurize cefls to 20 paig. Leak—test cells at 20 psig. Elvacum cells lnd pruge with nitrogen. Heat up fuel and coolant loops. Moisture reaching equitibrium in cells, First cell lsak rate at vacuum. Fill coolant system. Filf fuel systém with flush salt. Sample coolant salt, Sample flush nit. Drain flush salt. Prapare for fuel salt fill. VFIII fuel system with fuel salt. Sampie fuel ult.l Preparation for criticality afid powor operation. Criticality and taking the reacter to full power. Fig. 4.6 Return to Full-Power Operation (Typical Schedule) 8% - 59 . 9A-1 — Restart Equlpment After a Power Outage involved startlng the dlesels, restarting equlpment and resetting tripped instruments. ' 5J-1 — Return to Power After a Load and Rod Seram involved msnipu- ‘lating the rods and heat~removal system components to return to power. If ‘no instrument or equipment tests were necessary, thls ‘could- be done in less than an hour. | 11A, 5I, and 57 — Return to Power After a Drain is divided into 3 parts. Part (1) assumes a drain to both drain tanks followed by a transfer through the transfer lines, refill and return to power (thls is the design method of transfer). Part (2) assumes a drain to both drain tanks followed by a transfer through the fill lines, refill, and return to power. (This method of transfer was used to save time. Jumpering of interlocks was necessary.) Paft (3)'assumes a drain to only one drain tank followed by a fill and return to power. h 4.6 Reactor Startup After a Major Shutdown ‘ Each startup was dlfferent, requiring scheduling.between the com- pletion of maintenance and start of operation. The cell membrane was usually installed shortly'after the final closure of the primary.system. Assuming this, and that maintenance on out-of-cell components was com- - pleted by the time they were needed, a somewhat typical startup‘would be as shown in Fig. L4-6. 4.5 Changes Made in the Plant The number and klnds of changes made in a plant after it beglns ioperatlon are influenced by two different things: original design and ":changlhg activities. 1In some_51ngle-purpose systems the number of changes may simply reflect how well'the eriginal design met its goals. In an ex- perimental plant sfich as'the}MSRE “most of the changes may be required for experimental purposes or changes in the mode of operatlon. Ih any event, '_tlme spent in meking changes affects the availability of the plant. | Durlng the course of the MSBE operation many changes were made, but only a few caused significant deley in the program. 60 After the end of construction and the non-nuclear checkout of the MSRE, it was required that a change request be formally initiated, re- viewed and approved before any modification was made. (Section 13B of - rReférencg 22,7which prescribes the procedure, defines a modifigatioh as "a change in the physical plant which produces a significently different characteristic or function in any component or system.") In the 55 months from June 1965 through December 1969, a total of 633 requests for changes in the reactor system were initiasted, of which 512 were_approved. For the chemical processing plant, 113 change requests were initiated of which 87 were epproved. _ ' 7 | Table h—lQrsummarizes for each 6-month period the nufiber of reqfiests initiated and approved fbr the reactor. The table also categorizes the approved changes as to the reason for making the change and the type of change that it was. As indicated, most of the changes were aimed at filling the needs of normal 0peration and over half were changes in either instru- mentetion or setpoints. | | Table 4-20 surmarizes the change requests for the chemical processing | ci; plant. ‘ ' ' Although we know how many changes were made, we cannot accurately sum up the times required for meking the chenges. We can say, hc#ever, that in. - the case of the reactor change requests, the vast majority either took very little time to execute or were made while other work was going on. Exceptions include the work on the fuel offgas system in the spring of 1966. In the -processing plant, the summer of 1968 was spent in teéting and modifications, particularly of the fluorine disposal syétefi. Summary of Reactor System Requests Table L4-19 61 Type Change I3Y30 10 29 muafiompom 12 1k 12 10 75 uo e jusundysuy 10 | 61 60 34 25 13 234 TeoTI309TH 17 46 Bupdid 7 14 18 100 TeoTUBYOSN 11 11 11 21 moleq Reason for Change Requests 13430 19 serdureg TeTo9dg 14 anofiw#omxfi 26 90U TUIAUOCY uoigexadp TBWION 13 - 86 10k 59 25 11 12 12 373 TouuoOsIad puy Juswd by JO UO0F30930Id 12 37 Aq9383 ‘5 14 " Number of pasoxddy gy 137 T6 25 30 23 7 21 512 - Requests huonwvfivmm Ea 126 | 115 173 103 69 3h 32 ‘18 21 20 633 Period JTeH | 'lsfa ond 1st 2nd 2nd 1st 2nd 1st 2nd Te9x 1965 1966 1967 | 1st 1968 1960 Total After institution of change request procedure on June 21, 1965. a Summary of Chem Plant Change Requests Table L4-20 62 Type Change I5Y30 sjutodjag 12 UOTFBIUSTNI SUT 16 33 TBoTX309TH Jurdig 17 10 37 T8O TUBLUISK " nofeT Reason for Change Requests 13430 soTdueg TeIoadg - guswixadxy ~ 90U TUSAUO) uotgerady uotjexadp TewION b2 25 '89 TouuOsIag puy jusud by Jo uotT19930ad £q93eg Number of pasoxddy k- 25 gl Requests _wwummzdum 56 30 113 JTBH 1st 2nd 1st 2nd lst 2nd 1st 2nd 1st 2nd Period JIB3X 1965 1966 1967 1968 1969‘ Total' 63 4.6 Tabnlation'of:ReéordeduVariables athull power It seemed desirable to record a set of readings of all of the various | reactdr-variables'at normal conditions. However,‘even thbngh the MSRE ran ~ at full power much of the time, conditions were constantly changing due mainly to the varilous- experiments which were performed ‘Also some of the variables were not recorded regularly thus making it difficult to seleet the best time. After consideration of these factors, the 12 to 8 shift on 10/12/69 was selected as a period at full nOWet when conditions were fairly normal and considerable information was available. Values were taken from the various recorder charts, routine control room and building logs and the weekly, monthly, ‘and other check lists. These are tabulated in Tables 4.2l and 4.22. A snapsbot listing all of the computer‘ inputs was also retrieved from the computer tapes for 0400 on 10/12/69., These values are listed in Table 4.23. The location of the sensing element is deseribed briefly in the description column. The number and letters in tbe'identification column correepond to those used in the design drawings m and other design documents. ' They can be used to further\identify the variable. 64 Table 4,21, Tabulation of recorded variables A Identification o | Description _?g?i;?gg TE-100-A5 Line 100 (Reactor butlet) temperature | _120$°F_ TE—lOD—A6. Line 100 temperature | © 1169°F TE-lOl-ZA Line 101 temperature ; | 12005F TE-101-3A " Line 101 temperature | - 1215°F TE-102-3A Line 102 temperature | - o 1160°F TE~102-B4A Line 102 temperature o | | 11526F TE-102=-5A Line 102 (Reactor 1n1et) temperature | 1173°F TE-103-A1B Line 103 temperature ) | " 908°F TE-103-B1 Line 103 temperature | - 798°F TE-103-6 Line 103 temperature | | ~ 1150°F TE~-103-8 .; , Line 103 temperature o | | | 1223°F TE-103-B11 - Line 103 temperature — | | | | 983°F TE-103-14B Line 103 temperature '1200°F TE-104-5B Line 104 temperature : 918°F; | ' TE-106-5B Line 106 temperature 1080°F TE-108-7 Line 108 temperature _ 222°F TE-109-7 Line 109 temperature 227°F TE-~200-B7B Line 200 temperature 1018°F TE~200-D7B- Line 200 temperature | | 1025°F TE-200-A8E Line 200 temperature 1035°F TE-200-B8B Line 200 temperature ‘ 1010°F TE-200-C8B Line 200 temperature - 1015°F TE-200-A9B Line 200 temperature 1010°F TE-200-B9B Line 200 temperature . 1025°F TE-200-14A Line 200 temperature 1018°F TE-200-16B Line 200 temperature - 1015°F TE-200-19A Line 200 temperature ' 1023°F TE-200-AS~A1B Line 200 penetration temperature 965°F TE-200-AS-B1A Line 200 penetration temperature 642°F 65 Table 4.21. '(continued) Identifiqation Description. §87i§7§9 TE-200-AS-C1A Line 200 penetration temperature - 332°F TE-200-AS—-2A Line 200 penetration temperature 363°F TE-201-2B Line 201 temperature 1072°F TE-201-5A Line 201 temperature 1062°F TE-201~7B .. Line 201 temperature 1068°F TE-201-A9B Line 201 temperature 1078°F TE-201-B9B Line 201 tempé:ature : 1045°F TE-201-~A10B Line 201 temperature 1073°F TE-201-B10B Line 201 temperature 1059°F TE-201-C10B Line 201 temperature 1086°F - TE-201-A11B Line 201 temperature 1078°F TE-201-C11B Line 201 temperature 1083°F TE-201-D11B Line 201 temperature 1075°F TE-201-AS-A1B Line 201 penetration temperature 706°F TE-201-AS-B1A Line 201 penetration temperature 772°F TE-201-AS-C1B Line 201 penetration temperature 350°F TE-201-AS-2A Line 201 penetration temperature 350°F TdI-201A ‘Radiator AT (salt) 59°F Xpr 201 Radiator powerrrflk 7.1 MW FR=201 @olant salt flow 860 gpm TE-203-1 Line 203 temperature 870°F TE—203¥2 . - Line 203 temperature 107°F . TE-204-1A Line 204 temperature 1135°F TE-204-2A Line 204 temperature - 1155°F TE-204-3A Line 204 temperature 1205°F .. | TE-204-4A Line 204 temperature 1200°F TE-204-5A° - Line 204 temperature 1193°F TE-204-6A Line 204 temperature 1170°F TE-204-A7B Line 204 temperature 1178°F TE-204-B7B Line 204 temperature 1170°F TE-204-8B Line 204 temperature 1130°F 66 Table 4.21 (continued) Identification - Description '§g7i§7§9 TE~-206-1A Line 206 temperature - 1073°F TE~206-2A Line 206 temperature 1124°F TE~-206-3A Line 206 temperature | 1147°F TE-206~4A Line 206 temperature 1163°F TE~206-5B Line 206 temperature . 1150°F TE~-206-6A Line 206 temperature 1073°F PI-407 Leak detector pressure 100 psig RM-500 Cover-gas supply radiation 0.3 mr/hr PI-500A Helium header pressure 223 psig - FIC-500A Helium flow rate 6 liters/min. PI-500G2 Helium treating station pressure 250 psig PI-500M Reduced helium pressure 35 psig PI-501A Reduced helium header pressure 34.5 psig PI-510A 01l tank No. 2 pressure 7.5 psig PR-511D Coolant drain tank pressure 4.8 psig TE-512-1 Line 512 temperature 96°F FI-512A Coolant pump purge gas flow 0.6 liters/min PI-~513A 01l tank No. 1 pressure - 8 psig TE-516-1 .Line 516 temperature 107°F PR-516 Fuel pump pressufe 5 psig FI-516B ~ Fuel pump purge gas flow 2.4 liters/min, PR-522A Fuel pump pressure 5.2 peig TE-524-2 Line 524 temperature 93°F FI-524B Fuel pump upper off-gas flow rate 477 LI-524C 0il catch tank No. 1 level 147 FI-526C Coolant pump upper off-gas flow rate 0.04 liters/min. LI-526C 0il catch tank No. 2 level 18% RM-528 Coolant gas supply radiation 1.5 mr/hr PR-528A Coolant pump pressure 5.1 psig O 67 - Table 4,21 (continued) TE-702-1A Line 702 ;emperatufe Ident}figetion Description ‘§§7i;7§9 A 0,I 548 Helium oxygen-centent' 0.4 ppm A H 0. T 548 Helium moisfiuréftontent 2 ppm FI-548A Helium oxygen anelyzer flow 100 cc/min FI-548B Helium moisture analyzer flow 100 cc/min PdI-556A Main eharcoal bed AP | 3.4 psi RM-557 Charcoal beds off-gas radiation. 0.1 mr/hr . 'RM-565 -. Cell air radiation 1.5 mr/hr FI-566 Reactor cell ai:”oxygen.analyzer flow 100 cc¢/min A0z 1 566 Reactor cell aifaoxygen content 2.7% PR-572B Fuel drain tankrNoL 1 pressure 5.9 psig PR~574B - Fuel drain tank”No. 2 pressure 5.5 psig PR-576B Fuel flush tank pressure - 5.0 psig PI-589 Orerflow tank pressure : 5.1 psig' FI-589 Overflow tank bubbler flow rate . 27 psig PI-592 . Fuel pump pressure 5.4 psig FI1-592 Fuel pump bubbler flow rate_r- 25 psig FI-593 Fuel pump bubbler flow rate . 23 psig LR-593C Fuel pump level 53% FI-594 Coolant pump. bubbler flow rate 24.8 psig FI&SSSe:"' Coolang pump bubbler flow rate 25.5 psig ',LR-SQSC . Coolant pump,leve}r, 57% _ 7F14596 L Fuel pump bubblef flow rate - 25.2 psig RM-596 Fuel pump gas supply radiation 0.1 mr/hr 7Fi—598tf; : Coolant'pump“Bubbiefiflow'ratel 24.0 psig FI-599 . Overflow tank bubbler flow rate - 27.0 psig 1—599B5. Overflow tank level | 3 - 23% FI—600A~e Overflow tank bubbler flow rate - 26.5 psig - | LI-600B . Overflow tank level | 247 PI-701A - Fuel oil pump No. 1 pressure C 10 psig PI-702A Fuel oil pump No. 2 pressure 64 psig * 135°F AR it b e b e 68 Table 4.21 (continued) Identificétion Description --?g?i;?gg FI-703 Fuel pump lube o0il flow rate 3.8 gpm FI-704 Fuel pump coolant oil flow rate 8.2’gpm PI-751A Coolant oil pump No. 1 pressure 8 psig PI-752A Coolant oil pump No. 2 pressure 56 psig TE-752-1A Line 752 temperature 130°F FI-753 Coolant pump lube oil flow rate 4.0 gpm FI-754 Coolant pump coolant oil flow rate 6.7 gpm LI-806A Steam dome water level (FD-1) 0% LI-807A Steam dome water level (FD-2) 0% FI-810 Condenser water flow rate (FD-1) 40 gpm FI1-812 Condenser water flow rate (FD-2) 40 gpm FI-817 Offgas particle trap water flow rate 4 gpm TI-820-1 0T-1 water outlet temperature ° 87°F TI-821-1 O0T-1 water inlet temperature 80°F FI-821A 0T-1 water flow rate 8.4 gpm TI-822-1 0T-2 water outlet temperature 81°F TI-823-1 0T-2 water inlet temperature . 77°F FI-823A 0T-2 water flow rate 8.6 gpm TI-826 Treated water cooler (TW out) temperature 98°F RM-827 Process water radiation 3 mr/hr TI~829 Treated water cooler (TW in) temperature 108°F PI-829A Treated water pump pressure 70 psig FI-830 Fuel pump motor cooling water flow rate 4.6 gpm FI-832 Coolant pump motor cooling water flow rate 4.8 gpm FI-836A ~Drain tank cell space cooler water flow rate 63 gpm FI-838A Reactor cell space cooler No. 1 water flow rate 53 gpm FI-840A Reactor cell space cooler No. 2 water flow rate 59 gpm FI-844A Thermal shield water flow rate 50 gpm PI-851A Cooling tower water pump pressure 34 psig FI-851C Cooling tower water to cooler flow rate 273 gpm 69 Table 4.21 (continued) Identifieatidn Description ?g?i;?gg TI-854 Treated water cooler (CIW out) temperature 84°F TI-858 Cooling tower water temperature 80°F FI—859 - Thermal shield slide water flow rate 7 gpm FI-862A Coolant cell‘spaee'eooler (west) flow rate 20 gpm FI-864A Coolant cell space cooler (east) flow rate 20 gpm F1-873 CCP gas cooler water flow rate 15 gpm | FI1-875 CCP No. 1 and 2 oil cooler water flow rate 6.3 gpm . TI-881-2A Air compresser ‘cooler outlet water temperature 86°F TI-881~2B Air compresser head outlet water temperature 92°F ' PdI-900A CCP AP I o 8 psi | PI-927A Ventilation filter suction 2.2 in. H;0 PdI-927B2 Ventilation stack filters AP 3.3 in. H,0 PI-927C .VentilationISteEkifan suction 5.5 in. H,0 PAI-937A TR to 840 level AP ~ | 0.05 in. H0 PAI-938A SESA to TR AP 0.04 in. H,0 RM-6000-1 Reactor cell radiation 50,000 R/hr RM-6000-2 Reactor cell radiation 30,000 R/hr RM-6000-3 Reactor cell radiation 50,000 R/hr RM-6000-4 Drain tank celi'i-radietipn ' 10 R/hr RM#GGOb—5 Drain tank~celi redietion' 60 R/hr RM-6000-6 Drain tank cell radiation 500 R/hr . RM-6010 Coolant cell radiation 100 R/hr 'RRr8100 | Reactor power'e" 8.5 MW RR~8200 | Log reactor power 7 MW FI-9000 Instrument 'air flow - 20% PIC-9006-1 Emergency Ngrsetpbint7'” 65 psig FI-9006 - Instrument air flow (emergency header) 447 PdI—AfiZQAZHI Radiator air pressure drop | 677 ZI-AD2 Bypass damper position 10% FI-ADBAT_ "Radiator stack flow rate 70% TI-AD Radiator air inlet temperature 71°F TI-AD3-8A Radiator air outlet temperature 190°F L 70 Tahle 4,21 (continued) Identification .Description ' ?g?i;?gg PI-AR-1 Instrufient'air receiver tank pressure 88 psig TE-CC-1 . Coolant cell ambient temperature 88°F TE-CC-2 ” Coolant cell ambient temperature . 90°F TE-CC-3 Coolant cell ambient temperature 78°F TE-CC-4 Coolant cell ambient temperature 80°F TE-CC-5 Coolant cell ambient témperature 76°F TE~-CC-~6 Coolant cell ambient temperature 115°F TE-CC-7 Coolant .cell ambient temperature 105°F TE-CC-8 Coolant cell ambient temperature 103°F TE-CDT-1A Coolant drain tank top temperature 1195°F TE-CDT-3A Coolant drain tank top temperature. 1182°F TE-CDT-5A Coolant drain tank middle temperature 1200°F WR-CDT-C1 Coolant drain tank weight 3% Eil COP-1- - Coolant o0il pump No. 1 current. 0 amps Eil COP-2 Coolant oil pump No. 2 current 8.7 amps TE-CP-1B Coolant pump flange-neck temperature 245°F TE-CP-BZA' Coolant pump bowl-neck temperature " 786°F TE-CP-C2A Coolant pump bowl-neck temperature 795°F TE-CP-3B Coolant pump bowl top temperature 957°F TE-CP-4A Coolant pump bowl top temperature 965°F TE‘C?PSA Coolant,pump bowl top temperature . ' 965°F TE-CP-6A Coolant pump bowl bottom temperature 1042°F TE-CP-7A Coolant pump bowl bottom temperature | 1020°F TE-CP-8A - Coolant pump bowl flange top temperature 107°F TE-CPM-1B Coolant pump motor temperature 100°F SI-CP Coolant pump speed t 1780 rpm E1I-CP Coolant pump current 51 amps EwI-CP Coolant pump power | 56.5 kW LR-CPA Coolant pump level (float) . 4.8 in. TE-CPLE-A2 CP level element pot (lower) temperature 1140°F 71 ‘Table 4.21 (continued) Identification ~ Description §87i;7§9 TE-CPLE-A4 CP level element pot (upper) temperature 1097°F TE-CPLE-A5 CP level element pipe (top) temperature 1102°F TE-CPLE-A6 CP level element sensor (lower) temperature 515°F TE-CPLE-A7 CP level element-sensor (upper) temperature 372°F TE—CR~121V Coolant radiator inlet pipe temperature 1060°F VTEQCRQIZS. Flow venturi pipe temperature 1020°F LI-DC Decontamination- cell sump level 0 in. LI-DTC),T Drain tank cell sump level 0 in; | TE-FD-1-2A Fuel drain tank No. 1 (upper) temperature | 1115°F TE-FD1-13A Fuel drain tank No. 1 (middle) temperature 1170°F TE-FD1-16A Fuel drain tankwNo. 1-bayonet (top) temperature 1130°F TE-FD1-17A Fuel drain'tank No. l_bayonet (upper) temperature 1146°F TE—FD1418B ) - Fuel drain tank No. 1 bayonet (center) temperature 1158°F TE-FD1-19A Fuel drain tank-No.nl bayonet (lower) temperaturej lléidF TE~FD1-20A Fuel drain tank)No,ll bayonet (bottom) temperature 1161°F WR-FD1C Fuel drain tank No. 1 weight 0% TE-FD2-2A Fuel drain tank No. 2 (upper) temperature 1140°F TE-FD2-13A Fuel drain tank No. 2 (middle) temperature - 1150°F TE-FD2w16A Fuel drain tank,No. 2 bayonet (top) temperature 1115°F E—FD2—17A Fuel drain tank_No._2_bayonet.(upper)'temperature 1130°F TE—FDZ-lBB Fuel drainitank No;VZ ba&onet'(center) temperature 1140°F TE-FD2-19A Fuelldrainrtank No. 2 bayonet (lower) temperature 1140°F | TE-FD2-20A - Fuel drainltank No.(? bayonet (bottom) temperature 1140°F WR-FD2C Fuel drain tank No. 2 weight o - 5% TE—FF—lOO—Z Freeze flange 100 center temperature 885°F TE—FF—102-2_ Freeze flange 102 center temperature 974°F_ TE—FF—ZOO-Z Freeze flange,ZQQ-center temperature:- 754°F TE-FF—201-2 Freeze flange 201'center temperature 806°F TE—FFTeSA - Fuel flush tank (top) temperature 1132°F TE-FFTI-7B Fuel flush tank (middle) temperature 1174°F 72 Table 4.21 (continued) ‘ FT-201B-4A _Ideptifieation Desertpt;on_ ?g?g;?%g WR-FFT-C Fuel flush tank weight 69.6% PAI-FI-Al Vent. system roughing filters AP 0.6 in. E;0 PdI-FI-A2 Vent. system ebsolute filters AP 1. 6 in. HzO E{iI FOP-1 Fuel oil pump No, 1 current 0 amps EiT FOPQZ Fuel ofl pump No, 2 current 8.2 amps TE-FP-1B Fuel pump neck-flange temperature 305°F TE~FP-2B Fuel pump neck temperature - 559°F TE-FP-3B Fuel pump neck-bowl tempereture 962°F TE-FP-4B Fuei‘pufip wal top temperatpre 992°F TE-FP-5B Fuel pump neck-bowl temperature ~ 990°F TE-FP-6A Ffiei pfimp bowl top temperature 1030°F TE?FP+9B Fue1 pump neck-bofil temperature 955° TE-FP-lOB " Fuel pump Bowl_top temperature 1002°F TEwF?-liB Fuel pump flange top temperature 150°F TE-FP-12B Feel‘pump bowl center,temperatere'- 998°F TE-FPM-1B Fuel pump motor temperature - 120°F SI-FP Fuel pump speed N 1176 rpm Ei-FP Fuel pump current 44.5 amps Ew-FP Fuel pump power | 34 kW LI-FSC Fuel storage cell sump ievel | 1.1 in. FT-201A-1A Flow element 20l1A pipe temperature 1130°F FT“ZOlAFZA Flow element 201A top flahge temperature 1210°F FT-201A-3A Flow element 201A bottom flange temperature 1160°F FI-201A-4A Flow element 201A pipe temperature 1130°F FT-201A-5A Flow element 201A top flange"temperature 1210°F FI-201A-6A Flow element 201A bottom flange temperature '1250°F FT-201B-1A Flow element 201B pipe temperature '970°F FT-201B-2A Flow elefient”ZOIB top flange temperature | 1180°F FT-201B-3A Flow element 201B bottom flange temperature ~1170°F Flow element 201B pipe temperature 1130°F 73 s TE-FV-204-1B Freeze . shoulder temperature | Table 4,21 (continued)' Ideutifieation ;rbeseription arig?ié?ES FT-201B-5A Flow element 201B top flange temperature 1200°F FTfZOLB—GA Flow element 2018 bottom flange temperature !1130° “TE-FV-103-1B Freeze_val§e7103 shoulder temperature - 1000°F TE—FV—lOBeZA Freeze ralVep103 center temperature .390° TE-FVf103;SB ‘Freeze valve_losushoulder temperature _5@O° - TE-FV-104-3B Freeze valve 104 shoulder temperature 450°F _TE-FVf104-BA Freeze valve 104 adjacent’pipe,temperature 450°F TE~-FV-104-5B Freeze valve;lO& pot temperature o 590°F | TE~FV-104-6B Freeze valve 104 pipe temperature 650°F "__ TE—FV—105-2A Freeze'vaiue:idsécenter temperature . 1250°F_L' TE—FV-lOS:AAA Freeae.valvelOSjadjacent pipe temperature: 1160°F;1,- TE—FV<105?B4A Freeze valve-losmadjaeeut pipe temperature' 1190°F, TE-FV-105-5B Freeze valvejibsgpot‘temperature o o 1080° TE-FV~-105-6B Freeze valve 105 pipe,temperature.. 1120°F | TE—FV-lOSéZA Freeze valve 106;ceuter.temperature 1215'F TE-FV;106-A4A Freeze'value-106'adjaaent pipe temperature” 1140°F TE-FV-106-B4A Freeae.valve106nadjaeent pipe'temperature _ 1190°F TE-FV-106-5B Freeze valve ififipot temperature | | 1120° TE-FV-107-1A Freeze_valve.1Q7rehbu1der,temperaturep 500°F TE;FV?107;BB Freeze_value_;p7shouider temperature‘_‘_ 4909 TE—FVe107—A4 7 Freeze valve;;07madjaeeutpipe,temperature. S530°F 'TE-FV—107-SBr_ )Freeze'va}vei07"pottemperature; B 610° TE-FV-107-6A -FreeZevalve-io?.pot temperature'--” - 1540° . TE-FV=108-SB Freezemalye;iOBishoulder temperature,q:t,; 440°. ,VTE;FV-108-34 __FreEie,vaire ibé;adjacent pipe temperaturer ASOPV:gfl _rTEéFV;108;SB Freeze valve 108 pot temperature - 640°F TE—FVQiOB-GA ‘Freezemvalve ifiBepot temperature '5436?; TE-FV-109-1B ~ Freeze valve-lOQfshoulder temperature o 470°F | TE;FV-IOQfGA Freeze valve 109 pot temperature 595°F valve 204 7909? 74 Table 4.21 (continued§ Identification Description‘ ?fi?i;?gg TE~FV-204-24 Freeze valve 204 center temperature 235°F TE-FV-204-3B Freeze valve 204 shoulder temperature 830°F TE-FV-206-1B Freeze valve 206 shoulder temperature 710°F TE-FV-206-2A Freeze valve 206 center temperature 300°F TE-FV-206-3B Freeze valve 206 shoulder temperature 830°F TE-H 103 fieater 103 ;éfiperature | 908°f PI-HB-A Hiéh bay préésure‘ | - =0.19 in. H20 TE-HX-1A Heat exchahgércoolént out -temperature 1060°F TE-HX-4A Heat exchanger\cbolant in temperature 1180°F TE-HX-7A Héat-exchénéér shell (center) temperature 1165°F ZI-ID-A Inlet radiator door position 82.5% TE-OFT-1 Overflow tank pipe temperature 934°F TE-OFT-2B Overflow tank top temperature 1202°F TE~OFT-4 Overflow tank side temperature 1190°F ZI-0D-A Outlet radiator door position 79.5% LI~0T1A3 Fuel oil .supply tank level 647 RM-0T1 Fuel o0il supply tank radiation 1.7 mr/hr LT-0T2A3 Coolant o0il.supply tank level 58% RM-0T2 Coolant oil supply tank radiation 0.1 mr/hr TIC-0; R1-1 Helium oxygen removal No. l‘temperature 790°F TIC 0; R2-1 Helium oxygen removal No. 2 temperature . 1235°F TIC 0, R1-2 Helium oxygen removal No. 1 wall temperature 513°F TIC 0; R2-2 Helium oxygen removal No. 2 wall temperature 860°F TIC PE 1 Helium preheater No. 1 temperature 790°F TIC PH 2 Helium preheater No. 2 temperature 800°F TE PT-1 Particle trap temperature 360°F TE PT-2 Particle trap temperature " 360°F TE PT-3 Particle trap temperature - 360°F TE R-5A Reactor top temperature 1206°F TE R-6A Reactor top temperature 1210°F 75 \"Table 4.21 (continued) Iantification , | - Description | L v§g7i;7§9 TE R-7A ' Reactor neck'temperéture, - 799°F TE BréA Reactor neck temperature | | 680°F TE R-9 Reactor néék"temperatfire‘. o o | 600°F --.. TE R-10 | Reactbr‘neck temperature‘, - 534°F TE er? B Reactor side temperature l . 1190°F TE R-23A Reactor side temperature o | 1185°F TEKRrBZAV - Reactor bottom temperature . 1170°F TE R-34 . Reactor neck flénge temperature | 233°F TE R-35 - Reactor neck‘flange:tempe;ature 198°F TE R-36A | Reactor control rod No. 1 (upper) temperature 449°F TE R-37A ) | Reactor cqntrp%rbd No,lz (upper) temperature 431°F TE R-38A Reactor con;?blirbd‘Nq. 3 (upper) temperature 460°F TE R-39A Reactor cofitrol;bd,No. 17(1ofier) tem@g;ature 855°F TE RrQOA,E- Reactorlcontrol-rod No;HZI(lowér) temperature 786°F TE R-41A | ‘Reactor control rodNo; 3_(lower) tempergture 689°F TE R-43B Reactor graphite tubg”(;dfier)'fémperatfire- 1025°F TE R-44A Reactor neck (bottom) temperature o 1218°F TE R-46A ~ Reactor neck (upper) temperature | o ~ 250°F TE fif47 i . Reactor neck (uppe:)”;emperatfire | - 215°F TE'RtAS o Réactor neck (fiééér).temperature, _:__T: o 212°F " TE R-52 Reactor thermal well temperature | L ~ 810°F LI RC-C ‘Reactor cell sump level . - 0 in. '._PITRC=A ' Reactor cell pressure - =25 psig FI-1 Containment stack flow ~ 15% VRM—SIA_:f _:_ - Cdntainmeng_éfiéék;éléhé. ; S e 100 cpm ':BMngB. | Contaififiéflt?éfi#ck“beta gafima o 11 cpm CRM-SIC _COntéinméht'é;g@k;iog;né_ :3- o ' ” - 530 cpm r;LI;SC;A S 'Storage'éell.sfimp;igvel ' - 1.0 in. H0 ‘LI TC-A ) Spare cell.gfifip'iéfiél | S 0.4 in. Hz0 PI VI-1- Vapor suppreséiqn tank pressure 0.4 psig LI WI-A Waste tank level 107 in. H20 LI WTC-A Waste tank cell level | 2.6 in. H30 ~ Table 4. 76 21 (continued) l1dentification " Description ?g?i;?ég Main blower vibration <1 mil OACOT Official average coolant outlet temperature 1011.5fF OAFOT Official average fuel outlet temperature 1208.3°F Control rod No. 1 position 35.5 in. Control rod No. 2 position 44 in. - Control rod No. 3 position 43,1 in. Fission chamber No., 1 position 60 in. Fission chamber No. 2 position 65.7 in. Fission chamber No. 1 count rate 10* cps Fission chamber No. 2 count rate 10“cps Control rod No. 1 clutch current 143 ma Control rod No. 2 clutch current 144 ma " Control rod No. 3 clutch current - 147 ma Motor generator 2 current 28 amps Motor generator 3 current 32 amps Motor generator 2 voltage 52 volts Motor generator 3 voltage 52 volts Battery voltage ‘ 50.5 volts Motor generator No. 1 voltage 260 volts Motor generator No. 1 current 130 amps Inverter voltage 206 V Inverter current 89 amps Main blower No. 1 current v260 amps Main blower No. 3 current _ 280 amps Component coolant pump No. 1 current 0 amps Component coolant pump No. 2 current 88 amps Instrument air dryer purge rate 12 cfm Table 4,22, 77 (average of 3 phases) Tabulation of recorded heater current on 10/12/69 H203-1 werrio oS crpein B e St H-CR-1 15 H203-2 0 RCH-5 12, H-CR-2 17 H204-1A 3 RCH-6 15 H-CR-3 23 LE-CP-1 4 RCH-7 8 'H-CR-4 18 LE~CP-2 7 H102-2 13 H-CR-5 18 FV204-1 2.5 R-1 18 H-CR~6 19 FV204-2 1 R-2 19 H-CR-7 15 FV206-1 2 R-3 19 H-CR-8 25 FV206-1A 1.5 HX-1 0 200-13 17 H204-1 15 | HX-2 16 ' 201-12 13 H206-1 1 HX~3 .16 202-2 14 CDT-1 15 FP-1 6 200-14 6 CDT-2 11 FP-2 6 200-15 10 CDT-3 12 RAN-1 0 201-10 11 cp-1 13 RAN-2 0 201-11 6 cP-2 12 200-16 2 201-13 H200-1 10 201-14 1 202-1 7 | H200-11 12 102-1 3 204-2 10 _ H200-12 14 522 0 205-1 7 | H201-1 13 102-4 9 © 204-3 6 H201-2 10 102-5 1 . FT201A1 5 H201-9 16 103 26 FT201A3 5 . H100-1 2 FV-103 FT201A2 6 CRGH-L 16 H-104-1 8 - FT201Ar 4 RCH-2 13 FV-104-14 4 " FT-201BL 6 RGE-3 19 FV-105-1A 11 _ FT201B3 5 | ORCH-4 21 FFT-1 16 ' FT201B2 6 H100-2 18 FFT-2 15 FT201B4 6 H101-2 11 FD-1-1 18 0 H101-3 13 FD-1-2 17 ‘78 Table 4.22 (continued) - Description ?g?i;?gg Descripti9n_. ?;7%;729 | f Déécription » §g7i;7§9 FD-2-1 18 FV-104-~3 11 FV-108-3 3 FD~2-2 13 FV-104-4 10 FV-108-1 1 FV-104-1 | FV-105-2 ' FV-108-3 5 FV-104-3 FV-105-3 FV-109-1 3 H~-104-5 10 FV=104-7 13 FV-109-2 4 H-104-6 8 FV-106-2 8 FV-109-3 2 FV-105-1 12 FV-106-3 10 FV-109-1 Sl FV-105-3 11 FV-110-2 0 - FV-109-3 5 FV-105-1 7 FV-110-3 0 FV~110-1 0 FV-105-4 11 FV-107-1 T2 FV-106-1 11 FV-107-2 3 FV-106-3 9 FV-107-3 3 FV-106-1 6 FV-107-1 1 FV-106-4 FV-107-3 5 FV~106~1A 11 FV-108-1 3 FV-104-2 9 FV~-108-2 5 79 Table 4.23., Computer snapshot taken at 0400 on 10/12/69 228 Main charcoal bed pressure drop Scan Identification Seq. o : Description Reading No. ' 10/12/69 " EWM-CP-D 256 Coolant pump power 31.8 kW EWM-FP-D 255 Fuel pump power 34.8 kW FqI-569 349 Reactor cell evacuation flow 1.27rlit2r/min FT-AD3-A 40 Radiator stack flow | 195,000 cfm FT481?A 242 Containment stack flow 20,200 cfm FT-201-A 15 Coolant salt flow 849 gpm FT-201-B 28 Coolant salt flow 840 gpm FT—SIZ—A' 236 Coolant pump purge gas flow 0.65 liter/min FT-516-E 234 = Fuel pump purge"gas flow 2.39 1liter/min FT-524-B 235 Fuel pump upper offgas flow 133 cc/min FT-526-C 237 Coolant pump upper offgas flow 83 cc/min FT-703-A 238 Fuel pump lube oil flow ‘ 3.75 gpm | FT-704-A 239 Fuel pump coolant oil flow 8.04 gpm FT-‘753-A- 20 Coolant pump lube oil flow 3.87 gpm FT-754-A 241 Coolant pump coolant oil flow 6.54 gpm LE-CP-A 65 Coolant pump level 4.6 in. salt LT-0T-1-A 248 Fuel oil itank level 12.4 in. oil LT-0T-2-A 250 Coolant oil tank level 11.1 in. oil LT-524-C 251 0il catch tank No. 1 level 11.3 in. oil LT-526 252 01l catch tamk No, 2 level 16.0 in. oil LT-593-C ' 50 Fuel pump level 6.1 in. salt -LT—SQS;C-: 61 Coolant pump level 5.6 in. salt LT-596-B 54 Fuel pump level 5.1 in. salt LT-598-C 62 Coolant pump level 5.5 in. salt - LT-599-B 57 Overflow tank 1evel - 5.6 in. salt LT-GOO-B. 58 Overflow tank level 6.3 in. salt PDT—ADZ-A 24 Radiator air pressure drop 9.1 in. Hz0 PDT-556-A 3.4 psi 80 Table 4.23 (continued) Identification 222? ‘ Description Reading | No. c 10/12/69 PDT-960-A 230 Component coolant pump AP 8.0 psi PT-HB-A 233 High bay pressure -.07 in. H,0 PT~RC-A 348 Reactor cell pressure -1.98 psig PT-500-A 223 Helium header pressure 220 psig PT-510 27 0il tank No. 2 pressure 6.8 psig PT-511-C 225 Coolant drain-ténk pressure -- psig PT-511~D 218 Coolant drain tank pressure 5.7 psig PT-513-A 226 0il tank No. 1 pressure- 7.7 psig PT-516 347 Fuel pump pressure 5.0 psig PT-517-A 224 Drain tank suppiy pressure 8.5 psig PT-522-A 13 Fuel pump pressure 5.0 psig PT-528 66 Coolant pump pressuré 4,6 psig PT-572-B 219 Fuel drain tank No. l‘préssure 5.2 psig PT-574-B 220 Fuel drain tank No. 2 pressure 5.1 psig PT-576-B 221 Fuel flush tank pressure 5.3 psig PT-589-A 53 Overflow tank pressure 7.2 psig PT-592-8 33 Fuel pump pressure | 5.6 psig PT-608-B 222 Fuel storage tank préssure ~1.2 psig RE-NLC1-A 32 Reactor power 8.6 MW RE-NLC2-A 36 Reactor power 8.6 MW RE-0T-1-B 262 0il tank No. 1 radiation 1.8 mr/hr - RE-0T-2-B 263 01l tank No. 2 radiation 0.09 mr/hr RE-5C1-A1 9 Reactor Power 8.5 MW RE-S1A 277 Containment stack alpha 14% scale RE-S1B 278 Containment-stackbeta-gamfia 7Z scale RE-SIC 270 Containment stack iodine 21% scale RE-500-D 261 Cover.gas supply radiation 0.3 mr/hr VRE-SZS-B 275 Coolant gas supply radiation 1.5 mr/hr RE-528-C 276 Coolant gas supply radiation 2.1 mr/hr RE-557-A 273 Offgas from charcoal beds radiation 0.1 mr/hr 81 " Table 4.23 (continued) Scan ~ Identification Seq. . - ‘Description R . . Reading B : No. .o 10/12/69 RE-557-B 274 Offgas. from charcoal beds radiation 0.1 mr/hr RE-565-B - 271 © Cell air radiation L _ 1.0 mr/hr RE-565-C 272 Cell air radiation ‘ - 1.4 mr/hr RE-596fA' 280 Fuel pump gas supply radiation E | 0.1 mr/hr RE-596-B 282 Fuel pump gas supply radiation o | 0.1 mr/hr:. RE-596-C - 283 ~Fuel pump gas supply radiation- 0.1 mr/hr RE-675-A 284 Sampler cold offgas - : 1.1 mr/hr RE-675-B . 285 = Sampler cold offgas . : 4.2 mr/hr RE-678-C 286 Sampler hot offgas \ : . 2600 mr/hr RE-678-D 287 Sampler hot offgas - T 4000 mr/hr RE-827-A - - - 264 - Process water radiation o 25 mr/hr RE-827+B - 265 Process water radiation o 26 mr/hr RE-827-C 266 -iProcess]fiater radiation _ , 34 mr/hr RM-NCC1-A6 259 - Reactor count rate - 10,000 cps. RM=NCC2-A6 260-l Reactor:cfiunt,ratg _ : - ; 10,000 cps” RM-NCC1-A7 S Reaétorzpbwerf-- e S 9.6 MW RM-NCC2~A7 - 42 . Réactorgpower,-: . - 9.2 MW . - : RM-NCC1-A9 44 Reactor,§2riod - - - - -300 sec: - RM-NCC2-A9 .45 - Reactor period - .. - - ~150 sec .. SE~CP-Gl-A 52 ‘:Coolantipumfi'speed, SR L - 1775 rpm - SE-FP-El-A 11 Fuel pump speed . - - 1190 rpm .TEeADi—lA:. 184 - Radiatoisifilet air temperature . B7°F- | _TE-AD3¥4 185 fiRadiatofébutiét duct wall temperature 107°F TE-AD3-5A 186 Rédifitdr}ofitlet~duct wall temperature. - 120°F TE-AD3-6 = 187. Radiatdrgbutlet?duct.wall temperatfire, | 129°F TE-AD3-7A 188 - Radiétdrbfitlet duct,wall'tempefature 173°F. TE-ADS-SA__ . 190 Radiatbfogtletjair-temperature S © 178°F TE-CDT=2A 181 . Coolant drain tank bottom temperature ~ 1210°F TE-CDT-8 182 ' Coolant drain tank middle temperature 1203°F - TE-CP-A2B 127 Coolant pump bowl-neck temperature - | 804°F 82 Table 4.23,-(continuéd) 110 Freeze flange 101 middle temperature Identification 22:? Description ~ Reading . No. 10/12/69 TE-CP-1A 126 Coolant pump flange-neck temperature 247°F TE-CP-3A 128 Coolant pump bowl top temperature 955°F TE-CP-8B 125 Coolant pump flange top temperature 102°F TE~CP~9A 129 Coolant pump bowl middle temperature - 1018°F TE-CPM-1A 124 Coolant pump motor temperature | 96°F TE~CIS-D 90 Surveillance rig top zone temperaturei 1200°F TE-CTS-E 88 Surveillance rig mid zone temperature 1250°F TE=-CIS-F 87 Surveillance rig bottom zone temperature | 1220°F TE-DBE-1 267 Diesel house ambient temperature | 76°F TE-DL-1 137 Computer room ambient temperature 71°F TE-DL-2 138 Computér reference thermal plane temperature 69°F TE~-DIC-1 303 Drainffank-cellwambient temperature 148?F TE-DTC-2 304 Drain tank cell ambient temperature 144°F TE-DTC-3 305 Drain tank cell ambient temperature 151°F - TE~DTC-4 306 Drain tank cell ambient temperature 145°F TE-DTC-5 307 Drain tank cell ambient temperature 149°F TE~-DTC-6 309 Drain tank.-cell ambient temperature 150°F TE-FD1-1A 166 Fuel drain tank No. 1. top temperature 1073°F TE-FD1-3A 163 Fuel drain tank No. 1 bottom temperature '1149°F TE-FD1-12A 164 Fuel drain tank No. 1 middle temperature '1178°F TE-FD1-18A 167 Fuel drain tank No. 1 bayonet temperature 1158°F TE-FD2-1A 171 Fuel drain tank No. 2 top temperature | 1062°F TE~FD2~3A 168 Fuel -drain tank No. 2 bottom temperature 1130°F TE-FD2-12A 169 Fuel drain tank No. 2 middle temperature = 1159°F TE-FD2-18A 172 Fuel drain tank No. 2 bayonet temperature 1138°F TE-FF100-4 106 Freeze flange 100 inner temperature | 904°F TE-FF100-5 107 Freeze flange 100 middle temperature 648°F - TE-FF100-6 108 Freeze flange 100 outer temperature - 559°F TE~-FF101-4 109 Freeze flange 101 inner temperature 824°F TE-FF101-5 585°F 83 Table 4.23 (continued) Identification g;:? Description ‘Reading B No. | 10/12/69 TE—FF101-6 111 Freeze,flemgeplfil_buter temperature ' 454°F TE—FF102-4 112 Freeze flange 102 inner temperature 868°F - TE-FF-102-5 113 Freeze_tlange 102 middle temperature 631°F TE-FF-102-6 114 FreezetflangepIOZ outer temperature 541°F TE—FFZOOfA 115 Freeze flamge 200 inner temperature - 769°F TE-FF200-5 116 Freeze flange 200 middle temperature 546°F TE—FFZOO—G 117 Freeze f;ange_2000uter temperature 477°F ‘TE-FFZOi—4, 118 Freezefflamge 201 inner temperature 796°F TE-EF20175 120 Freeze flange_ZOl middle temperature 605°F TE—FF26146‘ 121 Freeze_flange_201 outer temperature 495°F 'TE-FFTélA 175 Fuel flush tank top temperature 1134°F TE-FFT-2A '173 .Fuel f;ush_tankrbottom temperature 1157°F TE-FFT—lO-V 174 Fuel il&sh'tank middle temperature 1174°F - TE-FP-1A 92 Fuel pumpneckeflange"temperature - 303°F TE-FP-2A 93 Fuel pump neck tempereturev 7 518°F TE;FP—3A; 94 Ffiei pmmp neck-bowl temperature 962?F TE-FP-4A 98 Fuel pump_bqwl_top'temperature 1004’F TE—FPQSA 95 Fuel pumprmeek-bowl,temperature :984°F, | TE~FP-7B 102 Fuel.pfimp~bqw1,;pwer temperature. {1212iF TE—FP-SBt..] -103 Fuel pump;béwlybdttpm.temperature,»_ 1208°F "TEeFP—QA:': 96 Fuel pumptfieek-bowl temperature - 950°F ' TEQFPQIOA 97 Fuel_pumpbowl_topetempereture 992°F __TEfFP?ilA" 100 ?Feelfpumpffiefige'top temperature 146°F TE-FPM-1A _104 fFuel-pfimpfmotor temperature | 117°F 7”TE-FST—10:' 217 Fuel storage ‘tank temperature S 81°F ':TE-FVIOB-ZB 39 _Ereeze_valve_103 center temperaturer 407°F VTE-FV104elB | 30=i'FreeZe_ve1ve104Lshoulder_temperature 462°F - TE-FVLDS%2B"* 47 _Freezeya;Yeplqszeenter temperetere_f”' 1236°F TE~FV106-2B 51 Freeze velre 106_center temperature 1214°F . TE-FV107-1B 144 Freeze valve 107 shoulder temperature 102°F 84 Table 4.23 (continued) Particle trap temperature Identifiéation ggz? Description Reading- No. 10/12/69 TE-FV107-2B 145 Freeze valve 107 center temperature 490°F TE-FV107-3B 146 Freeze valve 107 shoulder temperature 90°F TE-FV108-1B 147 Freeze valve 108 shoulder temperature 77°F TE~-FV108-2B 148 Freeze valve 108 center temperature 446°F TE-FV108-3B 149 Freeze valve 108 shoulder temperature 76°F TE-FV109-1B 150 Freeze valve 109 shoulder temperature 103°F TE-FV109-23 151 Freézéryalve 109 éentervtemperature’ ‘ 478°F' TE-FV109-38 152 TFreeze valve 109 shoulder temperature 108°F TE-FV110-1B 153 Freeze valve 110 shoulder temperature 106°F TE-FV110-2B 156 Freeze valve 110 center temperature 77°F TE~FV110-3B 155 Freeze valve 110 shoulder temperafure 88°F TE-FV111-1B 156 Freeze valve 111 shoulder temperature 88°F TE-FV111-2B 157 Freeze vfilve 111 center temperature 80°F TE-FV111-3B 158 Freeze valve 111 sfibfilder temperature 88°F TE-FV112-1B | 160 Freeze valve 112 shoulder temperature 89°F TE-FV112-2B 161 Freeze valve 112 center temperature 83°F TE-FV112-3B 162 Freeze valve 112 shoulder temperature 89°F TE-FV204-2B 17 Freeze valve 204 center temperature 254°F TE-FV206-2B 43 Freeze valve 206 center temperature 303°F TE-HB-1 176 High bay ambient temperature 83°F TE-HX-2A 34 HX fuel outlet nozzle temperature 1170°F TE-HX-3A 67 HX fuel inlet nozzle temperature 1207°F TE-HX-9A 91 HX shell center temperature 1185°F TE-MB1-1 134 Main blower No. 1 bearing temperature 86°F TE-MB1-2 133 Main blower No. 1 bearing temperature 68°F TE-MB3-1 132 Main blower No. 3,bearin§ temperature '93°F TE~MB3-2 131 Main blower No. 3 bearing temperature 68°F TE-NIP-2 345 Nuclear instrument penetrétion temperaffire 139°F TE-OFT~-6B 105 Overflow tank bottom temperature 1172°F TE-PT2YM 183 203°F -~ —— 85 . Teble 4.23 (continued) TE-SER-1 178 Speciel equipment room ambient temperature _Identification g::? b Description ““-" Reading o . No. ] R -.10/12/69 TE—PT-ZEMV 165 Perticie trap-temperature 167°F ° | TE-PT-2FF 170 | Particle trap temperature 94°F TE-R-2 | ~ 26 Reactor top:tempereture 1189°F TE-R-4A 73 Reactor toputempereture '1209°F~ .TE—RrISA" 78 Reactor side tempereture - liBi’F TE-R-18A | 79 Reactor side temperaturei 1196°F B :_TE-R:ZQA; 80 - Reactor side temperatureTb 1182°F o | IE&E525§" 82 Reactor side temperature .i,. 1178° ‘ .TE—REZQA, | 55 Reactor bottom temperature, 1182f TEferZAp_' | 83 Reactor“bottom tempeneture 1182°F TE;RrZBA - 84 | Reactor'bdttomtemperature' 1181°F . | _'TE-REZQA 56 ReactorrbOttem temperature 1180°f . TE-R-30A 85 Reactor bottom temperature 1181°F eTE-RrBlA,l 59 Reactor Bottom temperature ’ iiSOfF | TE—RrézAn 77 Reactor graphite tube temperature - é31fF’ " _TE;R,45Afp 76 Reactor neck upper temperature | 282°F‘._ TE-R-49 8 Reactor top temperature:, | 1185?Fp . .TE—R950'_'. 10 Reactorftop temperature | 11§1°F'Afl" TE-R-51 '_ 12 Reactor top temperature 1185°F _TE-RC§i£ o 290 Reaetor_cell embient temperature” '139°F TE-RC-2 291 Reector?eeiiiembient tempereture. 143°F TE-RC&Bp . 292 Reactor ¢éii afibient temperaturep 1495F . -TEfRC+4“:;' 294 Reectur cell ambient temperature:_-- _ 135°E | TE-RC-5 295 Reactor cell ambient temperature 138°F | TE;RC§6_;E | 296 Reactor cell ambient témperaturerw | - 127°F TE;RC#7p._ | 297 Reactor cell ambient temperature::,b-'":‘ | 1§2°F | TE-RC-8 - 298 V'Reeeter cell ambient temperature‘ft- ~ 149°F VTE-RC*Q | 299 . Reactor cell ambient temperaturebflf_ 139°F TE-RC-10 301 Reector cell ambient temperatureii:. 153°F 101°F 86 Table 4.23 (continued) Identification g::? | Description . : . Reading. _ . " No. , ' 10/12/69 TE-TRM-1 - ‘177 Transmitter room ambient températureé | ~ 81°F TE-VH-1 180 Vent house’gfib{ent temperature - 76°F TE-VI-1 143 Vapor tank .vi’ai.ter“ temperature | . 63°F TE-VI-2 | 289 Vapor tank aif temperatfiré _‘. 66°F TE-100-A1 5 Line 100 temperature | | 1207°F TE-100-A2 25 Line 100 temperature 1207°F TE-100-A3 46 Line 100 temperature 1208°F TE-100-1A 68 Line 100 temperature 1209°F TE-100-3A 70 Line 100 temperature 1208°F TE~101-2A 60 Line 101 temperature 1210°F TE-102-A4A 72 Line 102 temperature 1167°F TE-102-1A 71 Line 102 temperature 1165°F TE-102-5D 6 Line 102 temperature 1167°F TE-200-C7A 122 Line 200 temperature 1014°F TE-200-20A 64 Line 200 temperature 1022°F TE-201-A1B 22 Line 201 temperature 1068°F TE-201-A1C 20 Line 201 temperature 1068°F TE-201-A2B 18 Line 202 temperature iOll°F | TE-201-A2C 16 Line 202 temperature io10°F TE-201-B11B 123 Line 201 temperature 1068°F TE-201~1B 63 Line 201 temperature lo72°r TE~202-A1 7 Line 202 (well) temperature 997°F TE-202-B1 27 Line 202 (well) temperature 1007°F TE-202-D1 48 Line 202 (well) temperature 1007°F TE-522-1 135 Line 522 témggrature 84°F TE-524-1 136 LineA524 température : 101°F. TE-556-1A - 201 Line 556 temperature 94°F TE-702-1B 192 Line 702 temperature 132°F TE-705-1A 193 Line 705 temperature 142°F TE-707-1A 194 Line 707 tempefature: 140°F TE-752-1B 195-' tine 752-temperaturé' | 123°F! 87 » _Tab1g_4,23‘ (continued) ~ Cdmpensated ion chamber No. 1 position Identification gg:? Description L Rgadingg ent2 Mol ‘ 10/12/69_ TE-755-1A 196 Line 755 temperature 126°F - TE-575-1A 197 Line 575 temperature 127°F TE-791-1 140 Line 791 temperature . - 102°F TE-795-1 141 Line 795 t¢épé;éturé 146°F TE-804-1 215 Line 804 temperature 100°F - TE-805-1 216 Line 805 temperature 104°F TE-811-1 211 Line 811 temperature 81°F TE-813-1 212 Line 813 temperature 79°F TE-826-1 202 Line 826 temperature 99°F TE-831-1 206 = Line 831 temperature 102°F TE-833-1 213 Line 833 temperature 99°F TE-837-1 203 Line 837'temperature‘ 102°F TE-841—1 205 Line 841 femperature 106°F TE-845-1 207 Line 845 temperature 124°F TE-846-1 204 Line 846 temperature 107°F TE-851-1 210 Line 851 temperature 79°F | TE-874-1 208 Line 874 temperature 118°F TE-876-1 214 Line 876 temperature 106°F TE-916 200 Line 916 temperature 335°F TE-917 198 Line 917 temperature o 126°F TE-922° 199 Line 922 temperature | 118°F WM-CDT 246 Coolant drain tank welght 152 1bs _WM&FDL 243 Fuelrdrain:tank:No;'1_Weight | 0 1bs WM-FD2 244 Fuel drain tank No. 2 weight 468 1bs WM-FFT 245 Fuel flush tank weight 8800 1bs wMEEST 8 247 ;Fué1 étor€ge.tank.wéight 0 1bs XPM-201 268 Reactor power. T.6Mi ZM-FC1 257 Fission chamber No. 1 position 61 in. ZMEFCZ_ 258 - Fission chamber No. 2 position 66 in, ZM-NCR1 19 36 4n. 88 Tdbié?4.23 \(continued) Radiator outlet door position 85 in. Idenfificatipn g::? Description '-'Réaéing No. C : 10/12/69 'ZM%NCRé 21 Compensated ion chamber No. 2 position éhhfh.‘ ZM-NCR3 23 Cofipensated ion chamber No. 3 position - 44 in, — ZT-1ID 37 Radiator inlet door position 89 ih. ZT-0D . 38 89 5. FUEL SYSTEM J. L. Crowley C. H. Gabbard R. H. Guymon J. K. Franzreb 5.1 - Descfiption The fuel—c1rculat1ng loop consisted of a graphlte—moderated reactor, "a centrifugal type fuel pump with an overflow tank, and a shell-and-tube "heat exchanger, all connected'by 5- 1nch Hastelloy-N piping. The normal 'hbperating temperature was about 1200°F and fuel or flush saltrwas'circulated at 1200 gpm. When the reactor was not:in\operation, the selt was drained to one or both of the fuel drain tanks or to the fuel flush tank. Inter- connecting'salt piping and‘freeze valves permit filling the reactor or transferrlng salt between tanks by manipulatlng valves in the hellum sup- ply and vent lines. 5.2 Purginngbisture and'Oxygen from the SyStem In the fall of 196k, before salt was charged into the drain tanks, oxygen and moisture were purged from the fuel c1rculat1ng system and the © drain tanks. ThlS was accompllshed'by pressurizing the system with helium to assure that it was lesk-tight followed by a combination of evacuation and purglng with dry hellum“before and during the heatup. Details of the . heatup are. glven in Chapter 17. The hellum'was 1ntroduced at the fuel - pump whlch was in operatlon to provide clrculatlon in the loop. The system wes vented or evacuated . at the normal fuel pump offgas line (518) or at the ' salt transfer line (llne 110) to the fuel processing system. The latter .provided a 1onger flow path and thus a more effectlve purge. - : Flgure 5-1 gives the sequence of 0perat10ns used. Pe&ks of moisture _;in the effluent gas were observed at sbout 250°F and again at about. 650°F. The system'was evacuated-after,eacn gf;these_mo;sture pesks.'.Further heat- . ing tb:1130°F'didqn6t'eause”eny'Significent addf%ional moisture releases. rLater -analyses of salt samples 1nd1cated that the purging had been very effective. H20 {ppm} FLOW (liters/min) 'TEMPERATURE (°F) 1000 800 QFF-GAS MOISTURE ORNL - DWG 65-4469 600 400 ® READINGS AT LINE 110 A READINGS AT LINE 510 200 OUT OF SERVICE 0 - o EVAC. TO 18-in, Hg - 88| E e Ty Dpeia EVAC, TO Yo - : 1 o} (FFT AND VAC. . z?l’T‘l REACTOR * Spsig W" S psig % Spsig ————‘l D! 25inHg OPENED) Hg | 10 . HELIUM INFLOW 5t ] " - ' - 1 5 o T 11, ! L 1250 1000 750 REACTOR VESSEL TEMPERATURE 500 250 0 20 22 24 2 28 30 1 .3 5 7 9 T 13 5 OCT 1964 NOV 1964 ? Fig. 5.1 Initial Removal of Moisture from Fuel System 06 “ times were deemed acceptable. 91 5.3 Fuel—Circulating-System Volume Calibration After the flush salt had been added to the drain tanks and transfers made to fill the freeze valves (see Section 5.11), the fuel circulating system was filled and operated for 8 days. It was then drained and filled several times to check the drain times and calibrate the -system. The cali- rbration was done by increasing the drain tank pressure in increments and recording the drain tank weight and the differential pressure: between the ~drain tank and the fuel pump.r This is plotted in Fig. 5-2. The system was purposely_overfilléd-to determine the position of the fuel-pump over- ~ flow line and to test the Overf16w-tank level indicators.. Overflow oc- curred at readings,on,the twoéfuel—pnmn bubblers of 89.5 and 91%Z. A total -0of 105 1b of salt. transferred;to the oVerflow tank prodnced a.reading on . the level instrument of 11, 5% or 4.2 inches of salt in acceptable agree- ‘ment with the calculated response. i:5.43 Drain Times The temperature'distributiOn in the drain valve (FV-103) was con- trolled so that it would thaw in-9 to 11 minutes after an emergency drain ' signal., (See Chapter 20 ) The time required thereafter for the salt to drain from the loop- depended on which valves were open in the salt and gas ~ lines. ‘Normally the freeze valves ‘to both drain tanks were kept thawed but for a while after a salt fill one valve would still be frozen. An 1_emergency drain signal_acted;to,thaw the freeze valves and to open valves in drain tank vent lines and'in'the equalizer lines between'the gas in the 'fuel pump and in the drain tanks but it was considered possible that one or more of these valves. might not open. Tests were conducted, therefore, | hto measure ‘drain- times for various conceivable combinations of valving. On a normal emergency drain the loop.drained completely in 9~11 min. - after FV-103 thawed, With the equalizer valves open but one freeze valve ':7kept'frozen, the drain time. wasraboutr30 minutes, regardless of whether or ,'not the vent valves were open. With the equalizers closed, one freeze ~ valve frozen and the vents open, the drain time was 41 min. These drain AP BETWEEN DRAIN TANK AND PUMP BOWL (PSI) 92 ORNL-DWG 73-597 - - | O 24 22 —SALT IN BOTTOM 20 J1__ OF PuMP BOWL 18 / 16 /d—SALT AT TOP OF REACTOR 14 i 12 / 10 / 8 Z—— SALT AT BOTTOM OF REACTOR 6 | ] | 0 2000 4000 6000 8000 10,000 POUNDS OF SALT TRANSFERRED Fig. 5.2 Fuel Circulating System Calibration 93 5,5 Mixing of Fuel and Flush Salts During 235y fuel‘operetion, the uranium éoncentration of the flush selt increased an average of about 215 ppm each time it was used aftér fuel salt operation; 'Duringfz33U'0peration, with:a'loWer uranium concen- tration in the fuel, theVécrresponding increase should have been only - 39 Pbm. The actual 1ncreases ‘during the three flushes after 233U fuel cir- culation were 36, 42, and 39 ppm. Using these figures, approximately L0 1bs of fuel salt mixed with the flush selt during each flushing opera- tion. (For more details on salt analysis and 1nterpretat10n of results, see. Reference 23.) 5.6 - Primary System Leak During operatien‘the-eexi-air activity was continuously monitored to detect any leaks from_thetprimarysystem. ‘No leaks were detected until after the final fuel salt drain in December, 1969. At this time the cell- air activity didlincrease which'indicated a leak. Subsequent tests showed< that the lesk was at or near one of the drain-tank freeze valves (FV-105). The activity was mostly xenon W1th some iodine, krypton, and noble metals. 'Four days after the first release there was less than 25 curies of xenon . and less than 50 mlllicuries of 1od1ne in the cell. This was released to the atmosphere W1thout exceedlng the release rate permltted by the MSRE ,_safety limits.? o R S o , B - The prdbabillty of the leak resulting from corrosion seems. remote. A ;,rrev1ew of the operatlon of the freeze valve does not. 1nd1cate any excessive '.jthermal stresses.- No abnormalities were found upon rev1ew1ng the . constructlon xhrays and other 1nspection reports. It was noted, however, that the weld _*between the air shroud and the salt plplng at the freeze valve was not - fspeclfled as a full penetratlon weld This led to the suspicion-that the | - leak may be a crack that started at this p01nt and was propagated.by stress """ cycllng. Determination of the exact locatlon and nature of the leak will be attempted during the post-operatlon examlnatlons., More informatlon ~on the prelimlnary evaluatlon of the leek is given in Reférence 25. S 94 5.T Operation The operation of the fuel system was satisfactory. Difficulties en- countered with the components.are described in the following sections. The rate of transfer of salt to the overflow tank was higher than expected. Details on this and the loop void fraction and xenon poisoning are given in Reference 18. 5.8 Fuel Pump and Overflow Tank 5.8.1 Description The fuel salt circulating pump was a centrifugael sump-type pump with an overhung impeller, developed at ORNL expressly for circulating molten fluoride salts of the type used st the MSRE.26 (See Fig. 5-3 and 5-4.) At the normal operating speed of 1189 rpm, it had an output of about 1250 gpfi at a 48,5-ft salt head. About 50 gpm of the pump output was circulated internally to the pump bowl vie & spray ring to promote the relesse of entrained or dissolved gaseoué fission products. The gas space in the pump bowl was purged with helium to sweep these to the offgas disposal system. The helium was introduced Jjust below. the shaft seal in the béaring housing. Most of the gas flowed downward through the lasbyrinth between the shaft and the shield block to prevent radioactive gas and salt mist from reaching the seal. The remainder flowed upward to prevent any 0il which leeked through the seal from getting into the pump bowl. 0il for lubri- cating‘the bearings and cooling the shield plug was recirculated by an external pumping system. Helium bubbler type instruments were used to measure the liquid level as a means of determining the inventory of salt in the fuel system. Smell capsules were lowered into the bowl to teke safiples for analysis or to add fuel salt. The meximum height of the liquid in the pump bowl was limited by the top of an overflow line (5-1/2 in. ebove the center line of the volute) which connected to aS:Sfft3 overflow tank located beneath the pump. Helium bub- bler type instruments'were also used for measuring the liquid level in the overflGW'tank ' Slnce the overflow line extended to the bottom of the over- flow tank c1051ng & valve in the overflow tank vent line forced the salt back to the fuel pump. (fi; N ‘_/\\ J@\ S s 7 IO 220220 _ s et SAMPLER ENRICHER LEVEL INDICATOR SHAFT COUPL! SHAFT SEAL { See Inset) LEAK LUBE OIL BALL BEARINGS {Face to Face) {Bock to Back) L_UBE OIL OUT SEAL 0IL LEAKAGE DRAIN LEAK DETECTOR_ {Out of Section) {See Inset) BUBBLE TYPE OPERATING LEVEL To Overflow' Tank Fig. 5.3 BALL BEARINGS ~, ORNL-LR-DWG-58043-B BE OIL BREATHER NG HOUSING - GAS PURGE IN - SHAFT SEAL (See Inset) SHIELD COOLANT PASSAGES (in Porallel With Lube 0il) SHIELD PLUG GAS PURGE OUT (See Inset) GAS FILLED EXPANSION SPACE STRIPPER (Spray Ring) SPRAY MSRE Fuel Pump c6 ORNL-DWG 69- 10172 96 SUCTION m [ SAMPLE CAPSULE CAGE Cross Section'ofiMSRE Fuel Pump-Shbfiing Flow‘Paths; Fig, 5.4 lation no signlflcant dlffieultles were encountered durlng hellum circula- 97 5.8.2 EBarly Operationrwrd The fuel pump had been loop—tested with molten salt for 100 hrs at 1200°F before it was installed at the MSRE in October, 1964, After instal- tion while the fuel system.was belng purged, during early flush salt opera- tion which started on January 12, 1965, or during the criticality experiments. A continuous but very slow, ‘accumulation of salt in the overflow tank was | observed throughout the early»operatlon. (See 5.8.9.) 5.8. 3 Examination of Fuel Pump after Cr1t1ca11ty Exper1ment27 The fuel-pump rotary element was removed for inspectlon in September 1965 | ~...at the end of Run 3. The pump ‘had been used for circulating helium for 1410 hours and it had'been fllled:with salt at the MSRE for 2120 hours, 1895 hours of which the salt was belng clrculated _ - The pump was generally in. good condition ‘and appeared ready to be used for full-power operation. The only dimen31ona1 change was a 0.006~in. grovwth of the pump tank bore diameter where the upper O-rlng mated with the pump tank. The most 51gnif1cant dlscovery was ev1dence of 8 small 011 leak through _the gasketed Joint at the catch basin for the lower Oll seal. This oil had o run down the surface of the. shield plug, where it had become coked by the “higher temperatures near the bottom. Some of the oil had reached the upper O—ring'grooye at thepbottom of ‘the shield plug and,had'become coked_ln the | v:groove5?but-none*appeared}tofhawepleaked past the O-ring during high- ' temperature operation;'3Sofie7fresh>oil was observed below the ring after “ the rotary element had been moved to the decontamlnatlon cell, but it was _believed that thls oil drlpped from the open llnes durlng the transfer to 'gfthe cell.r- ’ — Lo ‘ All durlng subsequent operatlons, considerable dlfflculty‘was en- mfi“f&ountered from plugging of the mein offgas system, due largely to decom- - position products of ‘0il that had leaked into the pump bowl. (See " Chepter 8.) Because of this trouble -the gasketed Joint was seal-welded ‘on the spare rotary element._ However, the spare never had to be 1nstalled at the MSRE. I A layer of flush salt about 3/8-inch deep containing about 40 in.3 was trapped and frozen on top of the labyrinth flange. Apparently the salt had drained through the 1/8-inch diameter-holes_ih the flange'ufitil surface tension effects balanced the hydrostatic head. This. layer can be seen in Flg. 5-5, which is a photograph taken during the 1nspect10n. This photograph also shows the contrast between the surfaces exposed to the salt, which 'were bright but not corroded, and those above the salt, which were distinctly darker. There was a coating of fine sait mist paerticles on the lower face of the shield plug and in an "O"-ring groove around the shield plug there were small amounts of flush and fuel salt that must have - been transported as mist. The pump was reinstalled using remote maintenance techniquee so that these techniques could be evaluated. Four universal joints on the flenge bolts that had been fofind'broken during the disassembly were repaired prior Fo the.reinstallation of the rotary element. These failures resulted from excessive bolt torqfie that had been used earlier to obtain an initial seal on the flenge. (It turned out that the jack screws had not been fully _backed off before the flange bolts were tightened.) S,B;h " Pump and Pipe Support Problem ~ A problem related to the overall fuel-pump installation became evident during the post-criticality shutdown in the fall of 1965. The fuel pump could move in the horizontal plane, but was fixed against vertical movement. The heat exchanger could move‘pprizontally; the heat exchanger support frame was fixed against vertical movement at the north end, but was mounted on spring supports at the south end. The south end of the heat exehanger,wes coupled to the fuel-pump bowl by & short length of S-in. Sched.-40 pipe. The heat exchanger was supported from below, and the pump bowl was supported frofi above. The connecting piping was supposed to move upward at the heat exchanger and downward at the pump when the system was heated. | Because of the physical arrangement of the piping and equipment, stress ranges in the piping due to thermal cycling were calculated to‘reech a; maximum of 20,000 psi, which is acceptable. Uncertainties eiisted in - critical parts of the heat exchanger, however, particfilarly in the nozzle 0 O Fig. 5.5 Fuel Pump Rotary Element 100 where estimates were as high as 125,000 psi during cyecling from 150 and 1300°F. The end of the exchanger toward the pump bowl was therefore mounted on springs. This should have reduced stresses in the nozzle to the range of 20,000 to TO, 000 psi. ~ Careful observations during a. heating cycle to 1200°F showed that the equipment did not move as expected, however. Because of the complex equip- ment eonfiguretion and the inevitable uncertainty of the celculations, it was decided to make strain gage measurements with the equipment cold and moting the piping by mechanical meeans for measured distances with measur- gble loaeds. The highest stresses were fbun& to be in the'filletrwhere'the nozzle was welded to the hesat exehanger head when & spring force of 2000 poundS'was exerted to raise the end of the heat exchanger 3/16 in. The measured stress in the fillet was 13,000 psi and was a factor of sbout 5 greater than the stress in the nearby piping. Dye—check'oftthe'nozzle to the head of the heat exchanger showed no indication of cracks. | The conclusion from these tests’was that the mounting arrangement was adequate to allow the system to go through more then the 50.thermai cycles required without a fatigue faiiure. The system was then put to use for power operation.28 5.8.5 Effect of Bubbler Flow Rates on Indicated Fuel-Pump Level The fuel-pump level was determined by bubbling helium.through.the salt and measuring the differential preSSure between this line and & reference line which connected to the gas space of the pump (see Fig. 5-6). The end of one of the bubbler dip tubes (596) was 1.87hk in. lower than the other (593). A common reference line {592) was used for the two bubblers. The level readout instrument had a full scale (0-100%) range of 10 inches of selt. The centerline of the volute'was_at 35%. 'Compensation'was_previded in the instrumentation for chenges of density between flush end fuel salt and for differences in the lengths of the bubbler tubes. L Since the d/p cells used to indicate level were located outside the reaetor,cell,_there ves some pressure drop in the lines Between these and the'pump.‘ The smount of pressure drop was dependent upon flows through the lines. Tests were made early in 1965 to estsblish the reletionship between these flow rates and the indicated levels. 101 ORNL-DWG 70-5193 - 593 ™1.636in. R | { —1.874 in. ---—-—VOLUTE ¢ — o , L . "’/////////77/’///// 7y /- ///W 4 o —DENSITY ZONE Fig. 5.6 Schematic Representation of Fuel Pump Bowl Level and ' Density Indicators ' : L - 102 - Data for the upper probe are,presented'in Fig. 5-T. The curves for the lower probe were similar. It can be seen that the indicated level de- creased*with increasing pump-bowl refereace leg flow and that the indicated level increased with increased flow into the dip tubes. ' The actual salt level was not changed during the tests. Normal operating conditions were set at 25-psig forepressure on all three of the bubbler flow elements, 5 psig in the fuel—pump cover gas. ' 5.8.6 Fuel-Pump Level Changes and Limitations Based on the recommendetions of the pump .development group at ORNL, the level alarm and interlocks were originally set as shown in Teble 5-1. The differences in the level at'which the fuel pump could be started (64%) end the level at which the pump would‘automatieally stop (55%) was necessary because the 1nd1cated level decreased 10 to 12% 1mmed1ately upon starting the pump. This was due to fllling the spray rlng and fountaan flow chamber, : The narrow dlfferences between the high and low’ alarm and control set- points caused considerable operational difficulties. Prlor to starting the pump after a fill ‘the average loop temperature could not be sccurately determined. Since the fuel~pump level changed about 12 L% per 100°F change in loop temperature, the selected fill point was not always satisfactory for operation. In which case the freeze valve had to be thawed and the system level adjusted. During operation the reactor outlet temperature was usually ~ held constant when the power was changed. Therefore, the bulk average tem- perature changed with power changes and this caused changeS'in_the fuel-pump level. In addition to this, experiments were run which required operation at different reactor outlet temperatures. During load and rod scrams the system cooled rapidly. ©Sometimes it was necessary to reheat the system before restarting the pump. The eperating levels were further restricted when it appeared that the offgas plugged more rapidly when the salt level . was gbove 60 to 65% and gas entrainment in the fuel loop seemed to increase below sgbout 50%. The problem was further compounded by the changing pump- bowl level due to salt transfer to the overflow tank.l8 LR-593C (%) S—36 IN POSITION 3 70 - 69 103 ORNL-DWG 73-598 74 73 72 7 7, 0T A" = 67 | - ‘ L 66‘.r___,.——"' | 65 : ' 16 18 20 - 22 - 24 26 28 30 FI-593 (psig) NOTES: , FI-592 and FI-593 flow rates were proportional to the difference between the pressure upstream of the flow restrictors and the pressure in the fuel pump. The fuel pump pressure was held constant at 5 psig during these tests. - LR-593C was read from the bdttom of the inked space which was about 1.5% in width. . Fig, 5.7 Effect of Bubbler Flow on Fuel Pump Level Indicators .ok ' 4T5% 104 - Table 5-1.. Original FP Level Interlocks Level ; +78% - 464% T ¥55% ' 4539 Action and Reeson for Interlock ~Stops fill, gives a temperature setback.and rod reverse to prevent overfilling of the fuel pump (overflow point is about 90%). . Annunclatlon Pump cannot be started below thls level. This is to prevent cavitation. ‘ Annunciation Pump will stop below this level. Again, this is to prevent cavitation. | . 105 Tests were therefore performed whereby the interlocks were bypassed and the pump was started and operated at lower levels. Since there appeared to be no caVitation'or;adverse”effeets_on the pump, the interlocks given in Teble 5-1 were changed to 8%, T5%, 55%, 40%, and 38% late in 1965. Normal level during operation was still maintained between 50% .and 60% due to above,considerations._ HoWever,frecovery after a load and rod scram was much easier. 5.8.7 Coolant Air £o Fuel Pump | Tn the des1gn of the MSRE it was calculated that the upper portion of the fuel-pump tank wouldrbejspbjeot to substantial heating from fission .. products in the gas spaoe-above-the salt. Since the useful life of the pump tank would be 1imited;BYTthermal—stress considerations at the junction of the volute_stpport cylinder,with the spherical top of the tank, close control was at‘first maintained over the temperatures in this region. Design studies had indicatedethat the maximum lifetime would result if the junction temperature were'kept about 100°F below the temperature on the tank surface\6-in. out from the junction. Component cooling "air" _(95% Nz) was prOV1ded to. malntain this temperature distribution. A secondary ~consideration in controlling the temperatures was a desire to keep as Vmuoh of the pump tank as p0551b1e above the liquidus temperature of the salt. o : e o = In operating the reactor, it would.be 1dea1 if a fixed flow rate of ~air over the pump tank would.provide a. satisfactory temperature distribution = for all conditions.' Early de51gn calculatlons 1nd1cated that this condition ;could'be met with an air flow of 200 cfm.29 ‘However, temperature measure- R fments on the pump—test 1oop and during the 1nitial heatup of the MSRE indi- 'fircated that only ebout 50. cfm'would.be required, and that the air would have ',,:to be turned off when the pump ‘tank was empty. To minimize the temperature effects when the coollng air'was turned on, . air flow during power operation‘was ‘to have been set at the minimm that 'pjwould give the de31red temperatures. It was found thet an air flow of 30 cfm .:fgave 8 satlsfactory temperature distribution at all power levels. In_order to achieve and control;thisxrelatively low flow rate, a new valve, having 106 & lower C, had to be substituted for the original valve. Figure 5-8 | shows the temperatures in the two regions of interest as a function of re- actor power level with that air flow. The variations in the individual temperatureS'were caused'by varietions in the pump—bowl level salt temperar ture and air flow. Both the 1ndlv1dual temperatures and the temperature differences increased linearly with power, as expected. It was felt during early MSRE operation that, although temperature differences would exceed - 100°F at full power, the reactor could be so operated with the 30-cfm air fIOW'w1thout significantly reducing the 1life of the pump tank. When the reactor was started up for Run 8 in September 1966, an unex- plained'shiff-downward in these temperatures was noted. 'Latef the cooling air to the pump tenk was turned off during the attempts to melt out the salt Plug in the 522 line, and although the temperatures on the pumpetank surface were higher than with the cooling air, the temperature gradient was less. Since the temperature distribution was as good or better than with the air cooling, the use of air cooling was discontinued. 5.8.8 Salt Transfer to the Overflow Tank Early operation of the reactor showed that by some unexplained mecha~ nism, salt gradually accumulated in the fuel pump overflow tank even when the salt level in the pump bowl was well below the overflow point. The transfer rate depended on salt level, and the transfer ceased when the level was ebout 3 in. below the overflow point. This situation existed until sbout April 1966, when transfer began to be observed at lower salt levels. The rate appeared to incresse gradually as time went on until it ieveled off in June and July at 0.57 1b of salt per hour, independent of salt level as far down'ae 4,7 in. below the overflow point. The change occurred at the time of the stepwise increase in power, but no mechanism - connecting the two has been identified. This rate of transfer continued through the 235y operatlon and requlred emptylng the overflow tank about three times per week, _ Prior to operation with 233U fuel salt, the flush salt which had been processed to remove the 235y was circulated for 42 hours. During this and the first 16 hours of fuel carrier salt circulastion, the transfer rate and Qfi; ORNL-DWG 66-14441 1200 1100 |- 1000 ' TEMPERATURE (°F) 900 1 ) i | A ' A a A o ) ® & * a . @ ® - Q0 o O | C)'——oio——-l Fig. 5.8 2 3 4 5 6 7T 8 POWER (Mw) " Variation of Fuel-Pump Tank Temperafures with Reactor Power. Cooling-air flow, 30 cfm. LOT 108 loop void fraction appeared to be unchanged from the previous periods. Two hours after the start of a 12-hour exposure of & beryllium rod in the pump bowl, the bubble fraction in the system increased from the normal 0.1 vol % to about 0.6 vol % and.the rate of transfer to the overflow tank increased from sbout 0.4 1b/hr to greater than 4 1b/hr. , During the remsinder of the 233y operation the transfer rate was high (up to T2 1b/hr) and varisble. The. results of the investigation of the overflow rate, the changlng bubble fraction, and subsequent power pertur- bations are given in Reference 18 5.8.9 Burps of'the Overflow Tank During early opersation of the MSRE, when the bvefflow_rate'was-very low and the need to push salt from theroverfldw tank back to the fuel-pump bowl was infrequent, the practice was to empty the overflow tank completely. The sudden pressurization of the fuel pump at the end of the burp'géve false level indications and stopped the pump. Also when power operation was started, it was noted that gaseoué fissionlproducts fiere being pushed out the o0il seal line (524) by the sudden pressurization. Therefore, the procedures were changed such that the overflow tank was not completely emptied of salt during operation. In Fébruary 1969, the main offgas line plugged to the polnt where it presented a 4-psi pressure drop to the normal 3.2 liter/m offgas flow. During & burp of the overflow tank, this plug blew out with the reactor at full power resulting in a complete burp of the overflow tank. On &t least ' three other occasions when the overflow tank was being emptied, the offgas plug blew out causing more salt to be transferred then plenned. 5.8.10 Varisble-Speed Drive for the Fuel Pump - Prior to February 1969 the fuel pump was always operated.w1th the normal 60-HZ power supply at -~ 1189 rpm. Then, in order to investigate the effect of fuel circulation rate on system'béha&ior (Bubble ingestion, xenon stripping, transfer to overflow tank, ‘blips),18 a varlable-speed motor- generator set was brought to the MSRE to supply. pover to the fuel pump during experiments. As described in Chap. 16, ¢onsiderable effort was 109 expended in modifying ahd repairing the M-G set to obtain satisfactory reliebility. The pump itself, however, operated with no difficulty for substantial per;ode at speeds:between 50% and_los% of normal. - 5.8.11 Flooding of the Pum;g Bowl with Flush Salt . At the end of Run T in. July 1966 the fuel 1oop was filled with flush salt to rinse out residual pockets of fuel salt and thus reduce the radi- atlonrlevels for the scheduled work in the reactor cell, As the salt wes overflowing intorthe overflow tenk to rinse it out, the level in.the flueh_tank was lowered too far_fihich exposed the bottom of the dip tube. A large bubble of helium'gee:et—a preesure of ebout'30 pSig entered the - bottom of the loop. As the bubble rose, its volume increased due to the 'decreesed pressure and the fuel-pump level increased faster than it could overflow to the overflow tank. Salttflooded-the reference bubbler, the ennulus around the shaft, the”sempler tube, and'the'main offgas line. Severel factors contributed to the accident. Interlocks normally'pre— :vented filling to the overflow 1ine.. These had been bypassed to allow flushlng of the overflow tank. The flush-salt 1nventory was marginal at ' normal operating temperatures (1200°F) The flush-tank temperatures were “1ow (ebout 1140°F) which made the salt more dense and thus there was an’ ’1nsuff1c1ent volume. ;5.8 12 Plugging-of the Offgas Line at the Fuel Pump - Intermittent problemS'were encountered‘with plugglng of the 1/2-in. : SChedrhO mein offges line et a p01nt Just beyond'where it left the top of " the fuel-pump bowl. In-the early years of operatlon, thls either melted .',1tself free when the reactor was brought to power, or 1t was periodically ;reamed out vhen the reactor was shut down, through the use of a mechanlcal, flexible rotating snake";5 In July 1969, a specially built heater assembly ",'consisting of two 1000~w formed calrods wes 1nstalled remotely around the 7' 522 offges 1ine between the top of the pump bowl gnd’ the gas coollng shroud" : of -the fuel pump This, together Wlth beck—bIOW1ng'w1th hellum, was suc- CESSfUI in clearing the line. Mbre details on the offgas plugglng prdblems are given in Chepter 8. 110 5.8.13 Conclusions and Recommendsations The fuel pump was used to circulate salt for 21,788 hours at tempera- tures near 1200°F with no perceptible change in performance and no failure. Some plugging of the offgas line was encountered. This should be con-~ sidered in future designs. Perhaps dual lines could be used with heaters to melt out plugs if they develop. Leakage of oil (1 to 2 cc/day) into the pump bowl contributed to the offgas plugging prdblem. This possibility was eliminated in the replacement rotary element by seal-welding a gasketed joint, but because the plugging problem was managesble, theISPare element - was never installed. " The transfer of salt td the overflow tank was an operating nuisance in that it had to be periodically transferred back to the pump. The Mark-IT pump (which has operated for over.13,000 hours in a test loop) has more height in the pump bowl, eliminéting the need for an overflow tank. 5.9 Primary Heat Exchanger The primary heat exchanger was a horizontal shell and U-tube tjpe. During early power operation the heat transfer capability #as'foundlto be considerably lower than expected. A reevaluation of the physical properties ‘showed that the thermal conductivity of both the fuel and coolant salts was sufficiently below the value used in design to account for the overestimate of the -overall coefficient of heat transfer. Table 5-2 shows a comparison .of the physical property data used in the original design fo the current data. The heat transfer coefflclents calculated by the conventional design procedures using these two sets of data are also shown. o The heat transfer did not ‘change during subsequent operation of the reactor end no other d&fficulties occurred. The measured overall heat trans- fer coefficients ranged from 646 to 675 with an average of 656 Btu/(hr-ft2 °F) for 8 measurements. These measurements were made on the basis of nominal full power at 8.0 Mv and a coolant salt flow of 850 gpm. If the actual coolant flow rate were 770 gpm, which would be the flow consistent w1th 8 power level of T.34 Mw, the measured overall coefficient would be 59h as compared to a calculated value of 599 Btu/(hr-ft2-°F). !The performance is covered in detail in References 30 and 31. 111 Teble 5-2. Phy51cal Properties of Fuel and Coolant Salts | Used in MSRE Heat Exchanger Des1gn and Evaluatlon Original - o Current " Fuel Coolant Fuel wCoqlant Thermal Conductivity, Btu/(hr-ft-°F) 2.75 3.5 0.832 0.659 Viscosity, 1b/(ft-br) 17.9 20.0 18.7 23.6 Density, b/ft3 . 15L.3 1200 1h1.2 - 123.1 Specific Heat, Btu/(lb-°F) 0.6 0.57 0.4735 0.57T7 - Film Coefficient, _Btu/(hr-ft2-°F) 3523 5643 19T - 1989 Overall Coefficient,'Btu/(hr-ft2-°F) | 1186 - 618 112 5.10 Reactor Vessel and Reactor Access Nozzle -5.10.1 General Description | Reactor Vessel and Core — The reactor vessel was a 5-ft-diameter by 8-ft-high cylinder which contained a 55-in.-dismeter graphite core. A cutaway drawing of the reactor vessel and core is shown in Figure 5-9. The vessel design pressure and temperature were 50 psig and 1300°F with an ellowable stress»of 2750 psi. _ _ , , ' , - The fuel salt entered the flow distributor where it was directed down- wvard around the circumference of the vessel. It flowed in a spiral path ‘through & l-in. annulus between the vessel wall and-the core can for cooling purposes. Anti-swirl vanes in the lower head of the vessel straightened the flow path before it entered the graphite moderator core. Fiow“paSSageS5 formed by grooves in the sides of the graphite moderstor bars;'directed the laminar salt flow to the upper head plenum. The salt flow then left the re- actor vessel through the side outlet of the-reactor access nozzle which is described in the following section. _ : L During operation the fuel salt was held in the circulating system by a freeze valve (FV-103) attached to the lower head of the reactor vessel, The drain line and freeze valve was a 1-1/2 in. Sch.-40 pipe which was _flattened for sbout 2 in. to give a flow cross section about 1/2-in. wide. Cooling air in a shroud surrounding the flattened 'section maintained & frozen salt plug for a "closed" valve. The 1-1/2 in. pipe extended 2-3/k in. into the lower head of the vessel and was covered with a hood for protection against sediments collecting on top of the frozen plug meking it inoperative. The 1-1/2 in. hoodéd drain thus would remove all but a small puddle of salt from the lower head even if there had begn heavy sedimentation. To provide for drainage of the remaining puddle, a 1/2-in. tube was mounted through the wall of the portion of the drain protruding inside the vessel and ex- tended dowvnward through thé freeze valve below. This, in effect, formed parallel freeze valves operated by the same controls. The reéctor core consisted of 617 full and fractional size graphite | elements 2 x 2 in. cross section and about 67 in. long. Salt flow channels were formed by machined half channels on each of four faceé of each element. o 113 ORNL-LR-DWG 6109TRIA FLEXIBLE CONDUIT TO CONTROL ROD DRIVES GRAPHITE SAMPLE ACCESS PORT S ACCESS PORT COOLING JACKETS -~ . FUEL OUTLET — & REACTOR ACCESS PORT SMALL GRAPHITE SAMPLES HOLD-DOWN ROD OUTLET STRAINER _ CORE ROD THIMBLES - LARGE GRAPHITE SAMPLES CORE CENTERING GRID FLOW DISTRIBUTOR VOLUTE GRAPHITE - MODERATOR STRINGER FUEL INLET -/ g | _ . | N ~~— CORE WALL COOLING ANNULUS REACTOR CORE CAN ~ 1. _ ‘ REACTOR VESSEL —i: ANTI-SWIRL VANES — o < MODERATOR 'VESSEL DRAIN LINE - SUPPORT GRID - . Fig. 5.9 MSRE Reactor Vessel 114 When not buo&ed.by being immersed in salt, the vertical graphite elements were supported by & lattice of graphite blocks which‘in turn are supported by a Hastelloy-N grid fastened to the bottom of the core can. The ‘core can was thus free to.expand dounward relative to its top attachment to the're— actor vessel while the graphite was free to expand upward relative to 1ts support at the lower end of the core can. The grephite moderator elements were restrained from floating out of the core by a HastelloynN rod through holes in the dowel section &t the bottom of each graphite element. In addition a'W1re pessing through the upper graph1te elements prevented the upper portlon of & broken element from floating out of the core. To prevent possible overheating in an otherwise stagnant -region, & ‘ small portion of salt entering the resactor was_dlverted 1nto.the.reglon Just above the core can support flange in the annulus}betueen the vessel and . the .core can. . Vessel wall temperatures in this region and also the lower head were monltored during operation for p0351b1e deviations. , At the center. of the core were located sample specimens, three control rod thimbles, and f1ve removable graphite elements. . The- sample specimens were located in one corner of & U-in. square with. the three control rods occupylng the other three corners. The five,removable elements then occupied. the remaining spaces between these four,positions. ‘The controls rods are described elsewhere in this report (Chapter 19). | ‘ The sampleispecimen assemblies mentioned sbove were made of various com- binations and arrangements of graphite andrmetal.‘.They were exposed to much the same salt velocity, temperature and nuclear flux. as the core matrix. These are described in general elsewhere 1n thls section. ' ) The reactor vessel was supported from thettop removable cover of the ‘thermal shield by twelve hanger—rod.assemblies. These hanger-rod.assemblles were pinned to lugs welded to the reactor vessel just above the.flos.dis— tributor. Thetsupport arrangement was such that the reactor,ressel could be considered\to be anchored at, the support lugs. The portion of the vessel below the lugs was free to expand downward with no restraint. Reactor Access Nozzle (RAN) — Attached to the upper head of the reactor vessel vas a hofln._long 10-in.-diam nozzle. rThe nozzle had & 5-in.-diam ' { side outlet for the leaving salt located about 10 in. above the reactor 115 vessel upper head. The remaining extension of‘the 10-in.-diam nozzle pro- vided a pocket of trapped gas during the fllling of the reactor. However, most of the volume of this exten51on was occupied by the nozzle plug which is a removable support for the,three control rod thimbles, the 2-1/2 in. graphite sample access pipe,iand'for]the discharge screen. - See Fig. 5-10. Conventional_leak—deteeted metalyoval-ring—Joint3flanges were used on both sample‘access'openings"to provide the necessary containment of the primary system. The radia1 clearance between .the removablegnozsle plug and the - nozzle was 1/8-in. at the top and 1/h-in. at the bottom to provide a tap- ered annulus. It was 1ntended that -salt be frozen 1n this tapered annulus providing. a salt seal to prevent molten salt from contacting the metal seal- ing surfaces. However, maintaining a frozen salt seal was found to be not possible. This is dlscussed;in more detail later. Cooling air was pro- ‘vided on both inside of the plugland outside of the lO—in.Vnozzle but only to the inside of the plug of ‘the 2-1/2 in. graphlte sample access. Heaters were provided to thaw any frozen salt remainlng in the 10-in. annulus after . 'a salt drain. For thaw1ng the 2-1/2 in. annulus the cooling tube was re-~ moved from the plug as part of the sampling procedure and temporarily re- placed with a metal sheathed heater. Some twenty thermocouples (not including spares) were 1nstalled at 'varlous locations on the RAN. to monitor temperatures of the nozzle, both plugs and. control rod thimbles. A thermocouple well attached to the , graphite—sample access plug extended into the. flowzng salt stream to indi- - cate reactor outlet temperature. o o ' - ~ Strainers were provided et the reactor outlet to prevent passage of ~ graphite chips lerger then 1/16 in. A strainer basket vas attached to the "”Wlower end of the 10-1n. nozzle‘plug and extended downward 1nto the upper ';-head reglon of the reactor vessel.@ “The three control rod thimbles and the graphlte sample assembly passed through the strainer basket. Since the five graphlte elements were not pinned as were the remaining moderator ele- 'ments, a cross-shaped extension of the basket assembly projected beneath n the basket to provide a hold—down ~ See Fig. 5-11. ‘The. core sample spec1mens “were removed and replaced only: when the re- actor was drained of salt and partially cooled. The spec1mens were removed 116 ORNL -DWG 68-5507 - (coon.me AlR 19% in. ST A COOLING AIR FUEL SALT e L L L 10-in. ANNULYUS — ~ 2Y5-in. ANNULUS —— Fig. 5.10 Reactor Access Nozzle Showing 10- and 2-1/2 in. Annuli 117 ORNL~ DWG 636475 PLUG - FUEL OUTLET ACCESS NOZZLE TOP HEAD OF REACTOR VESSEL STRAINER — (16-goge PLATE WITH ¥-in. - ~diam. HOLES ON J4-~in. CENTERS) CONTROL ROD THIMBLE- GRAPHITE SAMPLES 2-in. GRAPHITE - - :-MODERATOR Fig.75.11 fR¢gctot,F¢e1 Outlet Strainer 118 through the smaller of the two access nozzles described earlier. A stain- less steel standyipe was left attached to the small graphite-sample access nozzle via a bellows. The upper end of the standpipe was bolted to a liner set into the lower concrete roof plug. During normal operation the liner opening was closed with a concrete plug. When semples were to be removed, this plug was replacéd with a work shield which contained openings for tools, lights, etc. All joints fiere moderately leak-tight and the stand~- ~pipe was provided with a nitrogen purge and offgas connections. A Roots blower located in the service tunnel provided a lightly negative standpipe pressure to assure inward leaskage. : Core samples were takeh by removing the graphite sample access flange, withdrawing the sample into the standpipe, placing the samplé in a special carrier, inserting a replacement sample specimen into the core, and re- placing the access flange. 5.10.2 Hest Treatment of Reactor Vessel After the reactor. vessel was 1nstalled tests of the particular heat of Hastelloy~N used in the vessel showed that the closure weld between the top head and the flow distributor ring could have poor mechanical proper- ties in the as-welded condition. Theréfbre in the fall of.1965, the‘reac- tor vessel was heat-treated in place, using installed heaters, for 90 hours at 1400°F. 5.10.3 Reactor Access Nozzle Freeze Tests and Effect of Circulating Bubbles The main purpose for a frozen salt seal in the'annuli of the 10-in. and 2-1/2-in. diam. RAN plugs was to prevent contact of the sealing surfaces with salt. The freeze flanges used as piping dlsconnects have a similar function. The maein difference being thet the freeze flanges incorporated a radial freeze joint while the RAN employed a longitudinal freeze joint. During development on the Engineering Test Loop freeze Joint, it was noted that a frozen salt seal could not be maintained relisbly if in con- ,fact with molten salt., It was intended to operate the MSRE RAN freeze ‘joint with & ring of frozen salt above the normal molten—salt level. This would be accomplished by pressurizing the fuel system above normsl operat- iné pressure soon after filling the reactor with salt, Cooling air on the 119 RAN would freeze & ring of salt in the annuli befbre the pressure was low- ered to its normal value, leaV1ng a gas void between the frozen and liquid salt. The gas trapped in the RAN annuli was thus the principal barrier between the molten salt andjthe containment sealing flange. The ring of frozen salt wouid then“serve'only as & backup protection against sudden pressure increases fcrcing salt up into thelannuli. | The first attempts to form this backup seal in this manner on the MSRE RAN were unsuccessful due to 1nsuff1c1ent cooling air although the availa- ble flow met the requirements of the design calculatlons.46 The flow was permanently,increased from aboutj3'cfm:to’the§range of 15 to 20 cfm by modifying the control-valveitrim. ‘Even with the increased air flow, there was sufficient movement of salt in the annuli due to the turbulent flow below to prevent forming a good frozen-salt seal, However, gas trapped above. the ‘molten salt kept the liquid level well below the flange. The ‘liquid gas interface changed 1n height due to any change in volume of the gas pockets., During early operation, the interface was at least a foot ‘below the flange and the flange temperature was about 200°F During later ,operations, there were periods when the salt level was higher and the tem- perature of the flange approached 300°F. There were two mechanisms which caused gas to be transferred to and from this pocket. Gastasgtransferred from the pocket’aS“a_solute and added by entrapment of circulating”bubbles from the salt stream below, The equilibrium difference'between these two transfer rates determined the " galt level in the annuli and thus the temperature readings of the wall. - Any change in reactor operating condltions which disturbed the balance' rd:between these two transfer mechanisms would thus change the height of salt _ in the ennuli. It vas noted that the salt level in the RAN annuli would ‘be increased by any of the following changes 1n operating conditions fhlowering the system pressure (cover gas in the fuel pump) increa51ng cir- culating salt temperature or decrea51ng the amount ofjcirculating bubbles o (i.e. decreasing pump speed).,; . Examples of all three of these causes are shown in Figs. 5-12 and 5-=13. Please refer to Fig. 5—10 fbr location of the thermocouples used in these two graphs. 120 ORNL-DWG 73—599 1400 EACTOR OUTLET - - '-.---———-_ N 12m —— F — ___jPJ _—\ '\ . ) TE R—438 c 2. 1000 m . S U TE R—7B % o0 |17} a. £ 800 P 600 —J 400 5 10 & o 3 J gollll | 1 1 | . | 1 1| 3 4 56 7 8 8 10 11213 14 1516 17 18 19 2021 22~ DATE BEGINNING FEBRUARY 3, 1968 Fig. 5.12 Effect of Reactor Outlet Temperature and Fuel Pump Cover- Gas Pressure on Reactor-Access Nozzle Salt Level ' (as indicated by wall temperatures) 121 ORNL~DWG 73-600 1400 1200 - FUEL PUMP SPEED ——1_——— A G T S S TE R-43 " TEMPERATURE (°F) FUEL PUMP SPEED (rpm) TE R—7 400 0 5 10 15 2 TIME, MARCH 11, 1969 (hrs) . Fig. 5 13 Effect of Fuel Pump Speed (i.e. circulating bubbles) on 'Reactor-Access" Nozzle Salt Level (as 1ndicated by wall temperatures) : , \ 122 The effect of temperature and pressure can be seen in Fig. 5-12. At the beginning of the dbsérved time period, the salt level in the RAN was relatively low with the pump pressure higher than its normal 5 psig, After a pressure drop on February 6, the salt level .overcame the immediate effects of pressure and began to rise as indicated by the R-U43 thermocouple. Sev- eral days laster the reactor outlet temperature was changed in two small steps of 30° and 15° yet note the very large effect of almost 600°F on the R-T7 thermocouple due to the salt level increase. Later by raising the pump pressure from 3 to 5 psig the salt level is again lowered in the RAN annulus. The effect of a direct change in the émount of circulating bubbles cén be seen in Fig. 5-13. At the beginning of this pefiod the fuel-pump speed vas lover than normal and it can be seen that the RAN salt level was very high.' Other indications such gsthé reactivity balance led to the conclu- sion there were no bubbles circulating at this qdndition. When”the fuel- pump speed was increased, with no other change in operating temperatfire or pressure, the RAN salt level dropped very rapidiy.}.(the the abscissa in Fig. 5-13 is now hours instead of days.) Circulating bubbles were involved in more dramatic effects than tem- -perature‘changes in the RAN, of course. Some more important variables were involved such as the effects of circulating voids end xenon poisoning on the reactivity balance. The effects seen in the RAN temperatures only helped explain the cause of some,dther;events. One phenomenom noted in which it is thought the RAN trapped gas had a direct relationship is that of power perturbations in the MSRE in January, 1969. There was an unusually large amount of circulating bubbles at this time as verified by several indications including low RAN temperatures. Itris thought some gas, clinging to the core, was suddently swept from the core causing a momentary increase in reactivity. The triggering mechanism for release of gas clifiging in the core was presumabiy a8 very small pertur- bation in flow or pressure. An occasional release of a burst of gas from the RAN which was suddenly compressed in the pump could have heen the cause of such perturbations.r There were other indications to connect the RAN trapped gas with the power perturbétions.“7 123 5.10.4 Core Specimen Installation and Removal Throughout the operation of the MSRE with salt in the primary loop there was a sample array of one kind or another in the core. The arrays that were exposed between September, 1965 and June, 1969 were of the de- sign shown in Fig 5-14. The array that was in the core from the time of construction until August 1965 contained similar amounts of graphite and INOR-8 (to have the same nuclear reactivity effect) but differed in inter- nal configuration., In 1969, during the last fivermonths of operation, a .‘ different array, designed'tc stndy the effects ofksalt velocity on fission product deposition, was exposed in the core. A core specimen assembly of the type shown in Fig. 5 14 consisted of three separable stringers (designated R, L, and S). Whenever an assembly was removed from the core, it'was taken to a hot cell, the’stringers were taken out of the basket, and a new assembly was prepared, usually including one or two of the previously'exposed stringers. Sometimes the old basket was reused, sometimes not. The history of exposure of INOR-8 specimens in this kind of array is outlined in Fig. 5f15. The numbers indicate the heats of INOR-8 from which the:rods in-each stringer were made. The following section describes the experience with installation and removal of core specimen assemblies, nith emphasis on the procedures, tools, ~ and systems involved. Descrlpticns of the materials that were exposed and the results of their post-operation examinations are given in detail in references 32—36. ( - Pre-Power Array —-The specimen array that was in the core during the . prenuclear testing and nuclear startup experiments differed internally from the later surveillance assemblies, but externally_it was similar and its installation and remcval'emplqyed the equipment and procedures prcposed for later use. The original installation was in 1964, before salt had been cir- - ‘culated, and,-although"dcne.remotely, was closely cbserveg to determine needs for modificatiqns in the tools and procedures. The assembly was re- moved in August, lQGSrby theiremote procedure. (At that time the salt was slightly radioactive from the *®°U startup experiments.) Only minor diffi- culties were encountered with some of the tools and fixtures. ©INOR-8 ROD OF TENSILE SPECMENS~,__ | GRAPHITE (CGB) SPECIMENS—__ . " BINDING STRAP~_ o AT e Y LN o A ‘/’/&\\\\\3/ Fig., 5.14 MSRE Surveillance Facility Inside Reactor Vessel C \ ‘ C el _ STRINGER L. STRINGER R’ STRNGER'S' " CORE TEMPERATURE - HIGH _SALT IN CORE RUN NO, 5085 - 50814 5085 5065 5085 5081 67-5514 7320 5085 . 5081 21545. || e7-502° . 67-551 21554 || 67-504 7320 1 0 4 ICIT1 10 1546 17 18 19 20 | Fig;‘S.ls Outline of MSRE Core Surveillance Pfogramf ORNL-DWG 72-7495 6C1 126 While the core access was.open for the sample removal, a visual in- spection was made of the core with a 7/8-in.~diam. scope. The inspection revealed that pieces were broken from the horizontal graphite bar at the | bottom of the core that was supposed to support the sample'assembly; The broken piecesrwere recovered, using a vacuum cleaner. To circumvent this damage, the new sample assembly to be installed was modified to be sus— pended from above, using a special fitting installed in the strainer bas- ket . above the core. This fitting (the "basket lock" in Fig. 5-14) had spring fingers that locked into the strainer screen when it was. pushed into place. The. lower end of the sample basket was extended to reach the hole in the lower horizontal graphite bar to provide lateral support.. Array 1. Installation of the basket lock and the first standard ar- ary in the core in September, 1965 was uneventful. The array was removed in July, 1966 after about 1087 equivalent full-power hours.(EF?H) of opera- tion. When the array was examined in the hot cell, portions of the string- ers were found to have been damaged. Some obstruction’(thought to be salt in the annulus) had been encountered in 1ifting the assembly through the access nozzle. The damage was not caused by handling, however, but by con- traction during cooldown. When the core was drained some salt was trapped between the ends of graphite speeimens'where it froze and interfered with the differential contraction of the parallel graphite and metal columns during cooldown. As a result the graphite columns buckled. Array 2. Because of the damage, none of the three Stfingers from the first array were included in the second. The new stringers were modified to,preventrtrapping of sait, but otherwise the array was like the first, The second array was installed on Sept. 16, 1966, near the end of the 2- month shutdown to replace the main’blowers. The second array was removed in May, 1967. By this time the reactor had operated- 4510 EFPH and the samples were removed -sooner after the end of power operation, so the salt was more highly radioactive than before, resulting in troublesome contamination on the tools and in the standpipe. This slowed the operation somewhat and it was necessary to install a char- coal filter in the standpipe vent line to limit iodine release to the stack. Difficulty was encountered in obtaining a satisfactory seal on the reactor access flange, which required feplacement of four bolts. Hot-cell examina- tion of the array showed that the basket lock had pulledGOut of the strainer 127 screen and was stuck on the baSket; A new basket 1ock was therefore fabri- rcated and was installed in the screen., _ Array 3. The third array consisted of one new stringer and two stringers previously exposed in Array 2, so it was highly radioactive at the time of installation. No- particular difficulty was encountered in handling the array, however, ‘since the equipment and procedure were de- | .signed for this condition,, A satisfactory seal on the core access flange _was'not obtained on the first try, and inspection_showed~the O-ring gasket was dented -and scratched..'Therbolt.holes'were cleaned and retapped and - a ,'new~gold—plated?ga9ketVWaslinstalled. An'acceptably tight joint was then ‘obtained. = 'lf--f o | - o Array 3 waS'removed on April 2, 1968 after the conclusion ofrzssu operation (9005 EFPH). Some inconvenience resulted when the closure on - the transport containment liner refused to operate properly, but the re- -moval and transfer were safely accomplished without the" liner. While the array was out. of the core the flange was sealed with a rubber—gasketed blind flange to avold possible damage to. the permanent sealing surfaces.- Array 4. Onme of the old stringers and two new stringers made up -_Array 4, which-was installed uneventfully on April 18, 1968 | This array was removed in June 1969 at 11 555 EFPH. The removal was delayed for _.several days -due to problems with the removable heater that was required to melt the salt from the tapered annulus between the removable plug and the 2—1/2—in. access nozzle.f For the first time, flush salt was not cir- : culated just prior to the sample removal The usual measures were taken 'gqto control the radiation and contamination during the sample removal opera- :ytion. There was some contamination of the immediate work area (mostly :nobledmetal fission produets) during removal of tools, which required ,mopping.'y _ o . ' ' e Special Array. The final array was all new and quite different in- 'f}ternally from previous arrays.r Externally, however, it was practically rthe same and standard tools and procedures were used in its installation 5(Ju1y, 1969) and-removalg(December, 1969) _ Both operations were uneventful. 128 5.10.5 Radiation Heating Radiation-produced heat in the reactor vessel walls caused the outer surfaces of the vessel to be hotter than tfieladjacent-salt by;an amount that was proportional to the reactor power. Any deposition of solids would cause even higher surface temperatures. There were two locations in the reactor vessel where solids could tend to accumulate. These loca- tions were the lower head and the lugs located just above the inlet volute and which supported the core matrix. The differences between thelreactor vessel temperatures and the salt iniet temperature were carefully moni- tored during power operation so that this condition would be detected had it occurred. _ W , There was no indication of sedimentation during the operating life of the reactor. The temperature difference did increase with use of **°U fuel; this was expected dué to the higher neutron leakage associated with this fuel. | 5.10.6 Discussion and Recommendations The reactor vessel and core satisfactorily performed all functions ‘ for which it was designed. Originally the operating life of the fuel sys- tem was to be'determified by the number of thermal cycles on the freeze flanges} However, the design life of the freeze flanges was extended on the basis of development tests. The stress-rupture life of therreaqtor vessel was then reviewed to determine if the operatidn of the MSRE would be limited_by the possibility of stress-rupture cracking. The reevalua- tion indicated the reactor could be operated at least-another'year longer than its originally predicted 20,000 hours s%ress—rupture life.’°!s’ The reactor access nozzle also adequately performed all of its func- tions. It was obvious, however, that some amount of bubbles circulating with the salt was'neéessary.to replenish the gas inventofy of the RAN an- nuli and keep the salt level down to the desirable level. The RAN design was thus inadequate from the standpoint of prdviding a frozen salt-seal such as was the case with the freeze flanges. The»ifiability to freeze a 129 salt seal was the result of a larger than anticipated salt flow in the RAN annuli which increased the heat load beyond the capability of the cooling air-provlded., Additional thermocouples‘in»the area of the RAN would have allowed better definition of the salt level. If an access flange such as the RAN is to be used again in a molten salt system, there are three possible methsds of obtaining its main func- tion — that of preventing.contact of salt with ‘sealing surfaces: (1) Provide an external source of gas to the annulus. A salt level determination would also be'neeessary. _ (2) Provide an internal source of gas in the form of circulating bubbles which would constantly replenish the gas inventory in the annulus. ‘A trapped gas pocket at a higher pressure than the pump cover pressure will not remain indefinitely. . (3) ‘Reduce the access of flowing salt to the annulus by baffles, ‘labyrinth or such, so thatia frozen salt seal can be maintained.‘ If bubbles are to be purposely injected into the fuel salt for the removal of xenon and krypton, then item (2) appears to be the best solu- tion. The joint could be designed to remain essentially full of gas at all times. | 5.11 Fuel and Flush Salt Drain Tanks 5.11.1" Description , Two fuel saltrdrsin tanks (FDFIVand FD-2) were installed‘in-the drain - tank cell for the safe storage ‘of the fuel salt during shutdown periods. “?A third tank (FFT), also located in the drain-tank cell, was provided for ,storing the flush salt which was ‘used for cleaning up the fuel circulating system before and after maintenance. 130. A fuel drain tank is shown in Figure 5-16. The INOR-8 tank was 50 in. in diameter by 86 in. high, and had a volume of 80 ft3, sufficient to hold in non-critical geometry all the salt from the fuel ¢circulating system. The tank was provided with a cooling system designed to remove 100 kW of fission product decay heat. The cooling was accomplished by boiling water in 32 double contained bayonet tubes and thimbles in each of the tanks. The fuel flush tank was similar to the fuel drain tanks, except that it had no cooling system. Since the flush salt did not contain fissile material,ffhe only decay heating was from the small quantity of fission products that it removed from the fuel system during the flushing operations. The weight of salt in each of the three tanks was indicated by forced balance pneumatic weigh cells.. The weights were recorded on strip charts but could be read more accuratély from instélle&,mercury manometers. Two ‘conductivity type level probes (one near thé]bottom'of the tank and the other near the top elevation of the salt) were provided for use as reference points. ~S;11'2 - Calibration of the Drain Tank Weigh Cells Using Lead Wéights During early testing in the fall of 196k the weigh cells were cali- brated by loading the tank supporting rings with lead blllets. The drain tanks were at ambient temperature during this calibration. Drifts in-indi-i cated weight of up to 39 1lbs were'noted when no.changes were being made. Repairs of air leaks in the instruments caused shifts of up to 65 lbs. The weigh cells were further calibrated during the addltlon of flush and 235U fuel salt. (See below.) 5.11.3 Addition of Flush.Salt and.Further.Calibration..of Weigh Cells Later in 1964, thirty-six batches of flush salt.(v250 1bs per batch) were added from cans in a portsble furfiace directly into drain tank FD-2 througha'heated line and flanged dip tube attached atethe‘inspectibn flénge'oh the tank. Each batch of salt. added was weighed on accurately callbrated scales and the tare welght of the container was checked to ob- tain the net weight of salt’ added. All temperatures were maintained as - near 1200°F as possible. During additions some difficulty was encountered with salt freezing in sections of the addition line. On one occasion the 131 ORNL-LR-DWG 61719 INSPECTION, SAMPLER, AND LEVEL PROBE ACCESS STEAM DROME WATER DOWNCOMER INLETS CORRUGATED FLEXIBLE HOSE STEAM RISER BAYONET SUPPORT PLATE STRIP WOUND FLEXIBLE % ' =1 ¥\ — GAYONET SUPPORT PLATE HOSE WATER DOWNCOMER : —a i P4 GAS PRESSURIZATION AND VENT LINES FUEL SALT SYSTEM _FILL AND DRAIN LINE SUPPORT RING VORI ) BAYONET HEAT EXCHANGER THIMBLES (32) 32\ =i s \ N THIMBLE POSITIONING RINGS FUEL SALT SYSTEM _ FILL ANDDRAINLINE =/~ _ TANK FILL LINE - . 'Fig. 5.16 Fuel Salt Drain Tank 132 drain-tank vent line plugged. To locate and remove the plug, the vent line was heated with a tdrch, starting at the tank and working down the pipe until gas could be blown through the line into the drain tank. The plug blew loose while heating the line several feet from the ténk. The plug location, the relative low temperature required to remove the plug and some oil found in the nozzle to the inspection flange on the tank led to the conclusion that the restriction was csused by an oil residue. ' During the addition of the first two batches of salt, weigh cell read- ings were taken every five minutes to determine when the probe light came on, Then this salt was transferred thiough the transfer lines to FFT and FDél and back to FD-2 through the reactor fill lines. This operation left salt in all freeze valves which were then frozen to isolate the tanks for the first time. - | While the salt was in the FFT, two more cans of flush sélt were added to FD-2 to recheck the location of the lower'probe. In the two checks the probe light came on when 469.6 and 492 1bs hed been added to the tank. These figures corrésponaed to weigh cell readings of 419 and 316 1bs. Af- ter all of the flush salt (9230 1bs) hed been added to FD-2, it was trans- 'férred between the three drain tenks. By using the weigh cells on.both tanks involved dnring each transfer, the weigh cells on FD-1 and;FFT were calibrated. The indicated weight at the upper and lower probés on FD-1 and FD-2 were also noted. The lower probe of FFT was not functioning (See 2,15}, The weight between the probes was 7353 + 176 1bs for FD-1 and T545 + 191 1lbs for FD-2. _ From the sbove data and sdbsequent operation, it was concluded that the weigh cells are~inadequate for precise inventory work. In general, the longer the time required for a given operation, the larger the error introduced. Transfers of small.qnantities of salt over short time inter- vals appeared to give fairly good data. Using the tank weight at the lower probe light as a fixed reference was useful in preventing the transfer of too much sslt to the fuel system and as an indication of the afiount of salt remaining in the tank after a'fill.- The implication is not that the weigh cells were at fault. Pipe loading on the tanks caused by temperafiure changes of the tanks and adjoining pipes probably caused the varistions. Errors as high as 200 to 300 lbs have been noted during transfers. 133 5,11, 4 U-235 Fuel Salt Addition The carrier salt and larger additions of 235U enriching salt were ad- ded by the same general method used for adding the flush salt. Details of these additions and the 235y criticality experiment are given in Ref. LO. 5.11,5 Salt Transfers . Salt is transferred between drain tanks by pressurizing the supply tank and venting the receivér-tank. - Originally transfers were made only through'the 1/2-in. transfer lines (107-110). In the event of a drain to both drain tanks followed immediately by & refill of the circulating loop, . considerable time was spent freezing the fill freeze valves (105 and 106), thawing the . transfer freeZe"valves (108 and 109), refreezing them after the transfer and then rethawing the £i11 freeze valves (105 and 106). After careful” cons1derat10n procedures were modified to permit jumpering inter- locks and transferring through the fill lines. No difficulties were en- .countered, however, all of salt could not be transferred this way since the fill lines did not go completely to the bottom of the tanks. 5.11.6 After Heat Removal - On 2/17/64 and 2/2L4/6kL, tests were made of the heat-removal capacity ofrthe-steam domes. Flush salt uas put into FD-2 and the salt was heated to ¢1200°F. With 40 gallons of water in the feedwater tank (FWT-2), a supply valve (LCV-80T) was opened to admit this water to all 32 of the bayonet cooling tubes. The”water was refluxed for approximately two hours “withrcooling tower'water-flow7to:the drain-tank coudefiserlkept coustant at 4o gal/min. The inlet and ex1t temperatures of the coollng tower water " were monitored, as were the drain tank temperatures and feedwater tank levels. At equilibrium condltlons, FD-2 temperatures dropped linearly at - 'a rate of 86°F/hr. The heat—removal rate calculated from condenser cooling .,water flows and temperatures was 139 kW/hr. The heat-removal rate calcu- -~ lated from the temperature—decay rate of ¥FD-2 and the heat capac1ty of FD-2 ~ was 132 kW/hr. S o | e A second similar test wes made bj'oyening the other supply valve (ESV-807) and keeping LCV-80T closed. The flow through this valve was 134 ~marginal, the level in the FWT decreased slowly and the heat-removal rate vas only 110 kW/hr based on a water heat balance and 98 kW/hr from the salt temperature decay rate. , , These data showed that the bayonet coollng tubes were more than ade— quate to remove the design decay heat rate of 100 kW/hr (for IO—MW opera- tion). Since the reactor was run at only approximately 8 MW, the margin of safety was enhanced. Because of leaks through the seats . of the water supply valves, the reactor was operated some of the time with the feedwater tanks empty and procedures were written to manually valve water to these tanks if neces- sary to remove afterheat. When the salt was divided about equally between the two drain»tarks, &s would occur on an emergency drain, no heat removal was necessary. Turning off the drain tank heaters was sufficient to prevent excessive temperature increases. For instence, on November 2, 1970, the reactor was drained while operating at full powér. 4380 1bs of salt drained to FD-1 and 5130 1bs drained to FD-2. With the drain tank heaters turned off, FD-1 temperature increased from 1060 to 1155°F in sbout 6-1/2 hours and then started decreasing. FD-2 temperatures increased from 1075 to 1195°F in sbout '9-1/2 hours and then started decreasing. 5.11.7 Plugging of Helium Lines.at.the Top..of. the.Fuel.Drain-Tanks The helium supply, vent and equalizer lines enter the top of the fuel drain tenks via a common 1/2-in. Schedule-40pipe. In June 1969, this line on FD-2 became almost totally plugged. The plug first appeared when the main 4.2 liter/min He offgas flow was routed past the drain tanks due to plugging in the regular offgas routes. The plug was at least partially cleared by first heating the tank to 51275°F overnight and then backblowing with helium at 42 psig. This technique was not successful when FD-1 line plugged in August 1969. Due to this plug, fuel drained preferentially to FD-2 during a planned drain on November 2, 1969. During the last drain on December 12, 1969, the fuel was made to drain more equally by.deliberately. closing the vent from FD-2 for part of the time during the draining period. The result was that the totsal salt weights in FD-1 and FD-2 were within 5% of each other after the drain had been completed. 135 5.11,8 Loading 233U into FD-2 After the fuel salt had been processed to remove the 235U, the re- maining carrier salt was transferred from the fuel storage tank to the fuel drain tanks. The method of addlng 233y enrlchment salt was d1fferent than that used for adding the flush salt and 233y fuel salt The loadlng equlp— ment as shown in Fig. 5-17 was attached to the access flange of FD-2, Cans contalnlng v 7 kg of uranium were lowered 1nto FD-2 whlch contalned half of the molten carrier salt. When the salt had melted out of the can, it was withdrawn into the shield. Mlxlng was accompllshed.by transferrlng the salt between the two drain tanks. .The 233y addition and crltlcality ex- perlment is described in detail in Ref. Ll. 5.11.9 Heatup and Cooldown Rates. These are descrlbed under heaters in Sectlon 17. 5.11.10 Discussion and Recommendations The fuel and flush salt drain tanks have fuhctiqned vefy satisfac- torily. They were'very sluggishrtfiermally and therefere easy to control during operation. The afterheat removal syetem'ffinctioned satisfactorily when needed. The weigh cells were not completely satisfactory for inventory pur- poses. Piping stresses should be eliminated or other type instruments provided. o As with other offgas lines, the p0931b111ty of plugging should be con- sidered. Perhaps two lines could be used w1th heaters attached to melt out any plug which mlght develop. ' S 136 ORNL-DWG 68-967 Il ——TURF CARRIER GRAPHITE SAMPLING SHIELD PURGE GAS NN SUPPLY o ‘.w |“E‘E by P o TOOL EXTENSION SEALS ~ TURNTABLE AND l STORAGE WELLS Y CONTAINMENT ENCLOSURE AND STANDPIPE ASSEMBLY Fig. 5.17 Arrangement for Adding *°*U Enriching Salt to Fuel Drain Tank’ 137 6. COOLANT SYSTEM _rC. H. Gabbard J. K. Franzreb i M. Richardson 6.1 Description The coolant circulating sYstem consisted of a centrifugal-type coolant punp and an air-cooled radiator which were connected to the fuel heat ex- changer by 5-inch INOR-8 piping. The normal operating temperature was 1000 to 1100°F at full power. The coolant salt was circulated at 850 gpm. " Dur- ing shutdowns the saltdwas—drained-to the coolant drain tank. The piping had low points on each side of the radiator. Therefore, two drain lines ‘and freeze valves were required. The circulatiug system was filled by pressurizing the coolant drain tank with helium and venting the coolant ‘pump. '6.2t Purgihg,Moieture~and Oxygen from the Bystem~ -~ 3Before;salt»wa3wcharged;intouthe coolantdrain‘tank,oxygeniandmois; ture were purged from the coolant circulating system and the drain tank. This was accompllshed by first pressurlzlng the system to assure that it was 1eak—t1ght then evacuatlng, ‘and purglng w1th helium during the heatup. Details of the heatup are glven in Sectlon 17, The hellum was 1ntroduced at the coolant pump and - coolant draln tank and evacuated or vented through the offgas system, The coolant pump was operated to prov1de circulation in the loop.' Purglng contlnued untll the system'was at 1200°F and essenti- ally no more-m01sture was . belng released Later analy51s of coolant salt samples 1nd1cated that ‘the purglng had been very effectlve. 6.3 CoolantfiCirculating;SYStem,Calibratibn.anchrainuEime _'After the coolantosalt:had been added to the COOlant-drain tank (see Section 6.1) the coolant circulating system was filled and operated for 11 days. A test of the drain time indicated that complete draining could .:? be accomplished in- approxlmately 12 minutes after the freeze valves thawed. During the following £111 the c1rculat1ng system was calibrated The drain tank was pressurized in 1ncrements and the drain tank welght and the dif- ferential pressure between the drain tank and the coolant pump were re-’ corded. The system calibration curve is shown in Fig. 6-1. DIFFERENTIAL PRESSURE BETWEEN DRAIN TANK AND LOOP (psi) 25 15 10 138 ORNL-DWG 73—-601 PUMP BOWL ——1g54 £ —jse50 g o ol - = S . O Q ' Q RADIATOR > w —J8ss o L o - / m L . w o W - -l - w : —1840 & LEAVING MAIN HEAT EXCHANGER oz ZZENTERING MAIN HEAT EXCHANGER ’<>E | i | SALT TEMPERATURE ~1130°F DENSITY | ~124.2 Ib/F3 ; | —1 835 800 1600 2400 3200 4000 4800 5500 POUNDS OF SALT IN COOLANT LOOP Fig. 6.1 Coolant Circulating System Calibration 139 - 6.h Operation Other than the difficulties with the radiator described later, the coolant system 0perated gatisfactorily. No salt leaks were detected at any time. 6.5 ~Coolant Salt Circulating Pump 6. 5 1 Descrlptlon The coolant pump used ‘to circulate the L1F-BeF2 coolant salt was simi- lar to the fuel pump. It did not have‘a spray ring since fission product gas removel was not required. No overflow tank pr cooling shroud was in- stalled. At the operating speed of 1750 rpm, it had an»output'of 850 gpm at a T8-ft salt head. A small helium purge was introduced Just below the shaft seal in the bearing housing. Most of the gas flowed downward through the labyrinth between the shaft and the shield block to prevent salt mist from reachlng the seal. The remainder flowed upward to prevent any oil which leaked through the seal from getting into the pump bowl. 0il for _lubrlcatlng the bearings and -cooling the shleld plug was reclrculated by an externsal pumping system., Heljium bubblers and a float-type level instru- ment were used to measure the liquid level. (A float-type instrument was not instelled on.the fuel pump.) The coolant pump was located outside the reactor cell and could be directly maintained shortly after shutdown. 6.5.2 Operation The coolent pump ran véry well throughout the eperetien”ef the reactor, ”ffAs W1th the fuel pump, the narrow renge of 0perat1ng 1eve1s caused opere- f?tlonal dlfflculties. Prior to starting the pump after a f111 the average loop temperature could not be accurately determlned Slnce the coolant—~ pump ‘level changed sbout: T% per 100°F change in. loop temperature the se- 1ected £i11 p01nt was not always satlsfactory for operatlon. ~In which case ‘the freeze valves had to be thawed and the system level adjusted The nor- mal operating level was about 55% at full power (the level 1nstrument had a span of 10 inches for 100% and centerllne of the volute was at 35%). 140 When the pump was first started after a fill, the level would drop about 9% due to a gas pockét in the loop and ebout 4% due to the pumé fountain flov. | | | During load and rod scrams, the system cooled rapidly which decreased the level below the interlocks required for starting the pump. Without circulation there was a possibility of freezing the radiator. An emergency start switch was installed which ensbled the operator to bypass all inter- locks and start the pump if freezing of the radiator appearéd eminent. Based on the oil found in the coolant pump offgas lines, there was a small oil leak into the pump bowl. The offgas line at the pump bowl be- came'pluggéd occasionally and in November 1968, a clamshell heater,was in- stalled. This was used once and cléared the line Bufficiéntlyrto permit operation until shutdofin_Decémber 12, 1969. - The float level indicator was not used for operation. 6.5.3 Conclusions and Recommendations. The coolant pump was used to circulate salt for 26,026 hours at tem- peratures between 1000° and 1200°F, Some plugging of the offgas line was encountered. This should be con- sidered in future design. Leskage of o0il into the pump bowl ‘could be elimi- nated by seal-welding a gasketed joint as described for the fuel‘pump (Section 6.1). | | B 6.6 Radiator 6.6.1 Description The nuclear power generated by the reactor was finally dissipated by a coolant salt-to-air radiator. Salt flowed through the 120 radiator tubes at 850 gpm where it was cooled from about 1075 to 1015°F at full power. Approximately 200,000 cfm of air from the two stack fans flowed perpenQ dicular to the tubes and out the coolant stfick. In péssing fhrough-the radiator, the air was heated about 100°F above the ambient inlet'tempera- ture. The heat removal was;édjusted to match the nuclear power produced by raising or lowering the two radiator doois, adjusfing the bypass damper, or starting and stopping the main blowers. 141 The radiator tubes were 0.75 in. OD x 0.072 in. wall x 30 ft long, zee-shaped to fit into the insulated radiator enclosure. The body of the enclosure supported the tubes and the electric heaters. Two hundred twenty-five kW of heat were provided to prevent freezing of salt in the tubes in case of a sudden lees of nuclear power (see Fig. 6.2). The upstream and downétream faces of the radiator enclosure were equipped with doors that could move up and down in a "U" shaped track. As the doors moved downward 1nto the fully closed position, camming devices 'operated by the weight of the door forced them against the seals located on the face of the radiator. The doors were raised and lowered by means of cables driven by a single-motor.' Magnefiic clutches and brakes permitted independent operation'ef;eitherfldeor. In case of a load scram request or an eleetrical pdwer.outage, the brakes and clutches were deenergized and” the doors fell closed at a rate which was limited by the inertia of fly- wheels attached to the ends of the drive shafts through overrunning clutches (nT ft/sec). | 6.6.2 Early Tests and Difficuities[ Preliminary heatup and checkout of the radiator began in October 196k, The main difficulties were due to insufficient insulation and warping of the doors. As & result, the doors jammed and would not operate properly and there was too much heat less to reach operating temperatures. The tem- perature distribution wasep00r?and overheating of thermocouple and electri- cal leads was encountered. | After unsuccessful attempts to correct these difflcultles, the doors were temporarily sealed to allow completlon of the crltlcallty tests while new doors were being de51gned and built. I During the last half of 1965, extensive changes, tests, and remodifi- 'cations were made on the radlator assembly. New doors were installed which | prOV1ded more insulation fac1ng ‘the heat source and addltlonal expan51on joints to prevent warpage.reThe_ 'soft" seal gaskets were removed from the doors and relocated on_thefradrator enclosure. They were increased from 3/4 in. to 1-1/2 in. widerfer”better sealing. A modified "hard" seal was installed on the doors. The door guide channels were repositioned to ORNL~LR-DWG 85844R2 142 o , \ TR A il Ly B ) e Ay sl % ] ; 8 .._ Ao NN N T. < - - ,.-_b... Radiétor Coil and Enclosure Fig; 6.2 143 provide more clearance, Four cams were installed on each door to provide more positive seals when they were closed. Sheet metal door hoods were installed above the radiator to reduce the air leakage into the penthouse, Thetsynchro indicating device on the drive shaft had not provedrrelisble since it was possible for the door to jam and the shaft to continue to rum in the downward direction. End rollers were p051tloned to prevent side motion from jamming the doors. Additional upper limit switches and mechanlcel stops were added to prevent lifting the doors in the event of a limit switch fallure. Llftlng motor,overload switches were installed in conjunction with the stops. | | The new doors weighed about 2700 pounds each which required the ad- dition of two 20,000-pound capacity, l-in. stroke shock absorbers on each door. Although it was fecognized that with the additional weight, operation ~of the existing brakes'and clutches was marginal, no changes were made in the lifting mechanism. Adjustment of the door-lifting cables so that the doors hung straight did much to ease their operatioh ih the slides, ~ The sbove modificatidns'improved:the heat distribution of the radiator system. Two major difficulties were.then encountered; one was the exces- sive temperature rise in the pent house just above the radietor, the other - was maintaining a negetive pressure in the pent house with respect to the high—bay pressure, This was needed for beryllium containment. Efforts to correct thése'prbblems ificluded relocating the heater lead Junction boxes, pulllng the thermoplastlc wire (194°F) from the junction f,boxes to the heater leads comlng from the radlator spreadlng out the bead- ed heater leads to improve the cooling air 01rculatlon_through the W1r1ng, ‘and, modifying the exhaustrdhoting to direct the sweep of cool air across ' the wire troughs (¢500 cfm);':These efforts were successful in reducing “the temperature to,aboot 100¢F}just above the insulation on top of the radiator. Heat leakagé;aroofi&.the'dOOr hoods and radietof'eficlosure was - reduced by weldlng up cracks, addlng 1nsulatlon, and 1nsta111ng boots - around the door—llftlng cable openings in the door hoods. A 1-7t2 oross- ‘overduct was installed between ‘the door hoods to relleve the pressure on the inlet hood. The pent house pressure with respect.to_the high bay was aided by the addition of a duct from the pent house to the inlet side of 144 the south annulus blower. This reduced the pent house pressure to -0.1h4 to ~0.28 -inches of water without the main blowers. With both blowers in cperation, the pressure became slightly positive, - 6.6.3 Subsequent Operation and Modifications The systems were heated and the approach to full power was started | early in 1966. The radiator functloned sat1sfactor11y, however, due to heat 1osses through the door seals, ‘it was dlfflcult to heat the empty radiator to- an acceptable temperature distribution prior to a f111 The heat transfer cepability was found to be sbout 20% less than cal- ~ culated. The cause for this dlscrepancy has not been deflnltely established but may be due to the large surface-to-alr temperature difference which existed. This is described in detail in References 20 and 36. Cooiant salt was frozen in the radiator tubes on two occasions. The first occurred on May 19, 1966 when a building'pomer failure caused arload scram and loss of all eiectrical equipment. Two tubes of the radiator were partially frozen as indicated by the temperatures.reading about T5°F low. These temperatures recovered a few minutes after'the coolant flow was re- established 1nd1cating that the plugs had thawed. ,‘ The second incident occurred on June 27, 1966 when again the electri- cal system failed, this time causing g8 coolant salt drain because of low radiator outlet temperature, The coolant salt weigh cells indicated that gbout 200 1bs of salt were held up in the coolant system. OSince it was probable that the salt was frozen in the radiator which had cooled below the freezing point, the radiator was reheated and the system refilled When the pump started, the temperature scanners indlcatefl that 5 tubes were fro- zen and not circulating. These tubes thawed after several minutes of salt circulation without damage to the tubes. | B | To reduce the possibility of this happening again, the control system was revised_So that 8 rod scram'wouldfalso cause a load scram. The low- temperature setpoint for a load scram was increased from 900°F, which is only 60°F sbove the liquidus temperature, to 990°F.2 Other revieions were made so that the coolant:pump would not be tfirnedleff unless absolutely aeceesary. A radiator door scram test was eenducted'from full power, and 145 "these revisions were adequate to prevent freezing of the radiator as long as the coolant pump remained running. The average temperature of the fuel and coolant leveled out at about 1120°F. | | | ~On July 17, 1966, a sudden loss in power indicated that a main dblower had been lost. Main blower No. 1 had failed catastrophically and pieces of the aluminum rotor hub and blading had gone into the radiator duct. Af- ter the radiator had cooled, examination revealed that pieces of aluminum | had passed through the radiator tube bundle, several dented tubes were found, and several pieées bf,aluminum were stuck_to tubes. Metallographic éxamination determined there were no punctures in the tubing and the radi- ator was safe to operate after each tube was thoroughly cleaned. The éxact cause of the blower failure was never specifically détermined but examina- tion of numerous "old"-cracks,in.the biades and hub would indicate that these were the probable point of origin. Flying debris caused some damage to the heater lead wires, wire trays, ete. S | It was apparent from inspection of the "hard" sealing surface, which was mounted on the hot side of the radiator door, that a modification would bé required. The continuous strip metal seal was badly buckled and frac- tured due to exposure to the radiator heat. The roughness of this surface damaged the "soft" seal (two widths of 3/l-in. Johns-Mansville Thermocore #C=19T7 square asbestos'packing'strips mounted on the face of the radiator) so as to render this seal ineffective. A short evaluation program of six types of "hard" seals resulted in 7 the,selection.of a seal madé*up'of-2-1/h—in. long overlapping steel seg- .ments spaced l/32—in.}apért'which was installed on the new doors. The test further indicated that the "I" bar, to which the segments were plug-welded, bowed as much as 3/16 in. on the 3-ft-long test sectlon.i The "T" bars in | ~ the new doors were provided with expansion slots plus a ten51on bar mounted ' }'on the cold 51de of the door which was used for stralghtenlng the doors after heatlng. b o Performance of the radiator enclosure and door was without incident after the above modificatiéfiéffiérefcbmpleted for the balance of the MSRE . 0peratifig period. Heat'logsgs_and mechanical'opefatidn were adequate. Programmed maintenance on the lifting mechanism was perfofmed when the op- portunity presented itself, Miscellaneous small items, such as replacement 146 ~ of the "soft" insulation along the top of the outlet door, replacement of broken ceramic insulators on the heater leads within the enclosure, repair of grounded heaters, etc., was the extent of the maintenance required. The door on the inlet side of the radiator retained its shape and ‘sealed reason- ably-fiell,'the seal along the top and sides of the outlet door retained its seal except for replacement of burned-out tape along the top. The bottom _ of-thé butlet door continued to distort ahd, even after being straightened several times, made & poor sesl as & result of the door bowing away from the soft seal on the radiator face, As long as the side and top seals were intact, there was no chimney effect and the heat losses were within ussable 6.6.4 Automastic Losd Control The‘heat removal rate at the MSRE was dependent upon the positions of the radiator doors and bypass damper and on the operation of the main blow- ers. These could be manually manifiulated as desired. Instrumentation was also provided to automatically sequence the operation of these-componefits so that the operator could use one switch for increasing or decreasing heat removal rate. An intermediate radiator door setting with one blower in operation and 1 MW of heat removed was to be used as the starting point. Since it was found that the heat removal with the doors at théir minimum setting of 12-in. was about 1.9 MW, the automatic system did not function properly. The rest of the sequencing functioned properly durihg a'test. However, in view of the above and since manusally adjfisting the heat load was not a disadvantage, the sutomatic load control was never used in the operation of the reactor. 6.6.5 Comments and Recommendations Assessment of the radiator containment, doors, and door-lifting mech- enism is complicated by the many changes, large and small, which were re- | quired before acceptable performance was achieved. o The radiator enclosure difficulties can be attributed prlncipally to details of installation such as gross air leakage into and around the en- closurg as a rgsult of cracks, unsealed openings, insufficient insulation, lgck of a sufficient number of expansion joints, ete. The prificip&l_radi- ator enclosure support "H" beam members were not vertical and therefore 147 the adjustable "U" shapped'dooriguides located within these members had ,to be remachlned to prevent the doors from binding., , The lifting mechanism was unduly complex. A substantial increase in both rellabillty and control would have been achieved by prov1d1ng each ~ door with its own motor-gear reducer unit with a built-in brake on the mo- tor rotor shaft. A small fast-acting brake in the high-speed end of the drive would have reduced the nuflber of electromechanical devices and re- -duced or eliminated the problem of coordinating time constants between ' clutches breakes, and motor;-'The ability to operate the doors indepen- dently was probably unnecessary The unrealistic close tolerances between the doors and the radiator enclosure end distortions produced by thermal gradients, primarily in the doors, required a massive effort to correct in order to,provide the re- : quired seal. These factorsjshould be given careful consideration in any u'future design. 6.7 Perfbrmance of Main Blowers, MB-1 and MB-3 6 7.1 Descrlgtion - The two main blowers are a part of the heat reJection equipment where the heat generated by the reactor is dlssipated to the atmosphere. The blowers were manufactured by the Joy Menufacturing Co. and are the "Axivane" type Model AR-600-360D-1225, The blowers are direct-connected to 250-hp, 1750~rpm motors by & short drlve shaft with a flexible shim - coupllng on - each end. _ : L o - . Although the blowers had been origlnally purchased for. another program, ~_ithey were essentlally unused at. the beglnnlng of the ‘MSRE program. The blowers were each orlginally rated at . 82,500 efm at 15 inches of water or '“illh 000 cfm of free air delivery.. The MSRE radiator design requirement of "100,00070£m-from each blowerrab 9 inches of water was'witbin ‘the normal L}_operating range'df'the bléwérs._ Had new blowers been procured for the iMSRE this same type of equlpment would have been considered & satis— "ffactory choice. L | : 148 6.7.2 Preoperational Testing - The pre0perationa1‘testing of the blowers consisted of sbout 25 hours of total operation., When the blowers were first deliveredénd installed, ‘they were operated & short time (less than 10 hours) to provide the elec- tricaliload for testing the diesel generators. The blowers were also oper- ated for a short period of time prior to power opération of the MSRE to check thé'radiatof tubes for self-excited vibrations, to.permit a calibra- tion of the stack air flow instrumentation, and td determine if control circuitry was necessary to avoid certain abnormal opérating conditions. During the abnormal operation testing, the blade-pitch was found to be less than the specified 20 degrees. The pifch was set corréctly and the test program was repeated. The tests indicated that the bloWerwaould operate satisfactorily at all the normal conditions, but that a "surge" condition would occur if the byrass damper were partialiy closed when the radiator doors were also closed. The manufacturer had stated during dis- cussions of the test program that the "surge" condition could be expected but that no damage would occur other than overloading the drive motors. The blowers were shut down immediately after the "surge" condition was en- countered and administrative control was used to prevent further operation under surge conditions. A second pitch change was made to 22-1/2 degrees immediately after the first full-power operation. This 22-1/2 degree pitch gave the maximum blower performance within thé power'rating 6f the drive motors. 6.7.3 Operating History The-aerodynamic performance of the blowers was satisfactory through- out the operating life of the reactor. However, there were several me- chanical failures which are listed in Teble 6.1 and which are discussed . below. This table also shows the effect of tne failure oh‘the~e§eration of the reactor. | - | | MB-1 Coupling Failure — After sbout 550 hduré of operation, the flexi- ble coupling on the motor end of the floating drive shaft failed during ' operation of MB-1l. The motor end of the shaft was then supported'bnly by the coupling guard, andrthe shaft whip that occurred as the blower rotor ‘?; ”TaB1e 6.1 'Summary of Main:Blower Failures Date ~Failure - Effect on Reactor Operation 6/14/66 Coupling Failure MB-1 Zero~Power Operation for 14 hours 7/17/66 ~ Hub and Blade Failure MB-1 Premature Shutdown from Run No. 7 3/6/67_' 1/16/68 ~ 8/31/68 9/17/69 9/19/69 ; Bearing:Fai1uréfMB~3 . . Bearing FfiilfiféfipMB-l'M3;3_ *"Be&ting Replééépent MB-3. Beafing and Mount Replacement MB-1 Faulty Bearing Replacement MB-1 ~ Partial Power‘Operation During Run No. 8 ' Zéro-Power.Operation'f@r 3 days " Partial Power Operation for 6 days. Maintenance Performed During Scheduled Zero-Power Operation ~ Completed During Scheduled Maintensnce ~ Period "Zero-Power Opération for 2 days Partial Power Operation 3 days Zero~Power Operation for 3 days YT 150 coasted down destroyed the guard and also the coupling on the end of the shaft near the blower. The shaft came to rest embedded in the sheet-metal nose with the blower end of the coupling near, but completely severed from, its mating flange on the rotor. The motor end of the shaft-ektended radi- ally across the blower inlet at about the L o'clock position. Scratches and dents on the blades indicated that some of the debris from the failed couplings had gone through the blowe} into the radiator duct. However, the damsage, except'to the couplings, appeared to be super- ficial and of little consequence. No additional inspecfionrof the blower was made. | » ' h The primary cause of the failure, wé beliéve, was a fatigue failure of the flexible shims. The-exact cause of the shim failurerisrunkndwn.‘ Some home-made washers were found in the drbrisrthat wére not roundedias they should have been. High stresses at the corners of these washers could have started the failure, or the coupling could have been running with excessive misalignment,- The blower rotor and drive shaft assembly had been rotated by hand when the final blade pitch change was made about 216 bperating- hours before the failure. There may have been some cracked shims in the coupling at that time, but the shims were still supporting the shaft. The couplings were rebuilt with new bolts, washers, and shims. The shaft was realigned, and the blower was returned to service. The couplings on MB-3 were also disassembled and new shims were installed.- The original shims in these couplings were not in bad condition. - | | - MB-1 Hub and Blade Failure.— After about 650 hours of additional op- eration, there was a second failure of MB-1l. This was a-éatastrophic fail- ure of the blading and the hub of MB-1. The outer periphery of the rotor hub disintegrated and all of the blading was deStroyed.'rfibst of the frag- ments wére contained in the blower casing, but numerous pieces of the cast aluminum-alloy hub gnd blades entered the radiator duct and some of these actually passed thrdugh but did not seriously damage the radiator'tube bundle., Figure 6.3 shows the failed hub and typical pieces of the blading. An inspection of the broken pieces of MB-1l revealed numerous "old" cracks in the blades and in the hub as evidenced by darkened or dirty areas on the fractured surfaces. One blade in particular had failed along a 151 ' "% PHOTO 84351 rFig. 6.3 Outer Periphery ofrFailed MB-1 Hub and Typical Blade Fractures 152 large "0ld" crack. The hub had contained short, 1-1/2 to 2-inch, circum- ferential cracks at the base of 8 of the 16 blade sockets and the failure generally followed these cracks. Figure 6.4 is a piece of the hub showing the darkened areas that indicated previous cracking at the basé of the blade sockets. Similar cracking was found in the hub of the spare blower “that had been in storage and a continuous crack extending dbout 35%vof the circumference as well as some of the shorter cracks were found in the MB-3 “hub. No cracking was found in the bladihg of MB-3 or the spare blower. The exact cafise of the MB-1 failure isruncertain, but a blade failure at - one of the existing cracks'éeemsrto be the‘most probable fifst,evént. All three of the blowers were rebuilt by the manufacturer with new re- designed hubs and lighter magnesium-alloy blading. The hubs were rein- forced with radial ribs to reduce the behding moments in the areas where the cracking had been observed. There have been no further cracking prob- lems in either of the rebuilt units in about 11,000 - 12,000 hdurs of op- eration. A more complete discussion of the failures and of the corrective action that was taken is given in References 42 and 43. Thrust Bearing Failures — There were a total of six failures assoc- iated with the main thrust bearing or its mounting. Main blower MB-3 was taken out of service on March 6, 1967 for a bearing replacement after the vibrations had increased from a normal value of about .8 to 2.5 mils. The bearing contained an excessive quantity of very old appearing grease, and the bearing surfaces were severely worn and pitted. This bearing was ap- parently from the replacement blower that had been in storage, and the bearing apparently hfid not been cleaned and relubricated when the rebuilt rotor was assembled. Improper lubrication was judged to be the cause of failure., The bearing had operated about 1800 hours. Main blower MB-3 was shut down a second time for a bearing replsacement on January 16, 1968, because of ;n increase in bearing vibration. Although the vibration and temperature had been normal, the MB-1l bearing was also found to be defective and was replaced at the same time. The MB-1l bearing had one pitted ball, but the MB-3 bearing was severely damaged and one ball had actually fractured. Figure 6.5 is a photograph of these two balls. Since improper lubrication did not seem to be the cause of these two failures and since three bearings had failed in relatively short operating Fig. 6.4 e £ST i 154 P;l;t 1 ™ |§ Y ¥ ‘.tij T BORATORY ” £ | OAK RIDGE NATIONAL LA Fig., 6.5 Ball Bearing Failures Taken from MSRE Main Blowers MB-1 and ' MB-3 in January 1968 ' 155 times, the expected'life of these bearings was-calculated from an estimated set of load conditions. The predicted life was 5000 hours as compared to operating times of 6630 and 4800 hours respectively forLMB—l and MB-3. As & result of this evaluatlon the original radial-type ball bearlngs were lreplaced with an angular—contact type having identical mountlng dimensions. The angular ‘contact bearing had a greater thrust load capacity. " The angular contect bearinQS'were'apparently capable of satisfactory 1life, but additional difficulty vas experienced with the self-aligning mount. The MB-3 bearing was replaced dur1ng a scheduled malntenance period lbecause of a thumping noise that wvas heard when the blower was turned by hand. The bearing was found to be in good condition, but the wear pattern indicated that the outer race'had been misaligned with the shaft. The mis- ‘alignment plus the high radial clearance in this type bearing csused the noise. No further dlfflculty was experienced with MB-3. | Blower MB-1 was shut down on September 17, 1969 because of an increase '_ in vibratlon. The vibration increase had been caused by excessive wear in the spherical surfaces of the self-aligning mount. The replacement of the mount also required replacement of the bearing. The new bearing failed by overheating during its_testlrun beceuserof insufficient clearnaces in the phenolic ball retainer. This bearing was then replaced with a bearing of the original radial type that was on hand. This bearing was satisfactory for the remaining operation of the reactor. The defective bearing was re- turned to the vendor. ~ 6.7.4 Surveillance of Blower Operation Following the'hub'andflblade failure of MB-1, a surveillance program 'f[for monltoring the operatlon of the blowers was 1n1t1ated Vibretion pick--‘ ups were mounted horlzontally on each bearlng of the blowers and drlve amotors and thermocouples were installed 1n each of the blower bearlngs. - A v1brat10n meter w1th fbur 1nput channels was prov1ded for each blower, 'and the v1brat1on of each bearlng was recorded in peak-to—peak dlsplacement once per shift by the operating personnel.. The thermocouples were moni- tored by the computer and 1n1tially an slarm would be given 1f the tempera- ture exceeded & preselected llmit. Thls was revised later to alarm if the 156 temperature rise above ambient exceeded a limit. This permltted much tighter operatlng llmlts because the ambient fluctuated with the outside air temperature. _ In addltion to the dlsplacement readings ‘taken from the V1brat10n ;meters, an output -was also available to display thevylbretlon trece on an oscilloscope or other high-speed reccrder, Oscillosccpe photographs were taken for reference on new bearings and-when the vibration meters indicated an increase. Figure 6,6 shows. a comparlson of vibration traces from a nor- mal bearlng and from the bearlng containing the fractured ball. A complete inspection of,the blowers including a dye-penetraht‘examina- ticu'cf the hubs and blading wes completed annually duringrshutdown periods. 6.7.5 Discussion and Conclusions Following the major rebuilding of the main blowers.in 1966, thefcper- ation has been generelly satisfactory even though there have been several power reductions caused by bearing problems. The total lost operating time caused by the variousbeering‘failures_was 8 days of zero-power operation and 9 days of partial-power operation. A better thrust bearing arrangement could have been required. | - - The surveillance program to monitor the mechanical'terformance'6f the blowers has been very effective. All the bearing problems except the MB-1 pitted ball in January 1968 have been found by an increase in vibration. When the ball fractured in the MB-3 bearing, a temperature‘ihcrease had preceded the vibration increase by several days. However, this-tempera- ture increase had gone unnoticed beceuse the bearing temperature was well below the limit that had been set. The computer program was therefbre re- vised to alarm on the temperature rise above ambient, and much tighter. limits were selected based on the previous operating experience. This alarm system was effective in detecting the defective new bearing on Sep- tenber 19, 1969. : The 0perat1ng experlence W1th the v1brat10n monitoring system has | clearly shown that the displacement meter alome is not adequate to mcn;tor the condition of the bearings. Some system of maintaining a history of the frequency spectrum is also requlred We used photographs of the oscil- loscope trace to document the normsal operatlng vibrations, and additional 157 ; . .- " S s iy i sl ik I Trace with Fractured Ball — Taken 1/16/68 Fig. 6.6 ' Main Blower MB-3 'Vib-rati:on Traces 158 photographs were taken for comparison when an increased vibration was indi- cated by the dispiacement meters.- A mbdification tortheieofiputer to in- crease its samplihg rate would have{permittednthe f:equeney spectra to be generated by the computer. This would have provided'a more convenient and more_preciée spectrum, but some degreerof:backgrofind experience would still be required for_properAinterprefation of the spectra in either case because there can be several normal frequencies. There #ere'several instances of abnormal vibrations when the bearlngs were not at fault. Tfie final de- cision to replace a bearing was made in each case by the sound and feel of the blower durlng a coastdown from operatlng speed. 6.7.6 Recommendatlons The main blowers were conventional commercial equipment which would normally be expected to operate with a minimumrof attention or maintenance. The instrumentation of all rotating equipment is'probably impractical in regard to both cost and manpower. However, certain classes of equipment should be instrumented and monitored on a routine basis. The equipment should be selected on the basis of: l. Size and cost 2. The'accessibility durihg operation 3. The effect of shutdown on overall plant operation i, The potential for self-destruction | Ideally the monitoring should be computerized but this is not actually required. _ , | Each type and piece of equipment will have its own operating characte- ristics in regard to vibration or temperature, and_these characteristics may vary with the insfalletion or operating conditibns.' Preliminary limits may be obtained from the manufacturer or from Reference hh but the final operating limits should be selected from a background of normal operating experience. ) The experiences W1th ‘the cracked hubs and with the orlglnal thrust bearing design are indications that the Quality Assurance Program should extend into the design phase as well as the fabrication phase of major or critical items of equipment. 159 6.8 Coolant Drain Tank 6. 8 1 Description The coolant drain tank vas - similar to but smaller ‘than the fuel flush _ tank. It was 40 in. in diameter by 78-in. high end hed & volume of 50 ft3. As there was essentially nofafterheet in the coolant salt, cooling was not - provided;_ Two selt lines (204 and 206) and their associated freeze valves were used for filling and draining the coolant circulating loop. ! - The weight of selt was indicated by forced balance weigh cells. The weights were recorded on,strip oharts but ‘could be read more accuretely from installed mercury manometers., Two conductivity type%prObes (one near the bottom and the other_near;the top elevation of the salt) were provided for use as reference points. 6.8.2 Calibration of the Weigh Cells Using Lead Weights . . The weigh cells were calibrated by loading the drain tank with lead billets. The drain.tank was at ambient temperature during the calibration. There was good agreement between the manometer read1ngs and the actual weight added.. Difficulty was encountered with air and mercury leaks in the readout system. The weigh cells were,further calibrated during the addition of coolant salt. - d | 6.8.3 Addition of Coolant Salt: Twenty-two batches of coolant salt (5,756 1bs) were added from cans in & portable furnace directly to the coolant drain tank through a heated line attached to the top of the tank. After the addition was complete, this B line:vas cut off gnd,welde&fgfifit;"fAVgood:linear plot of the manometer f readings vs aCtual fieight,was”obtained_except for a 150-1b step in the '5eurve'fihich‘occurred wheh-a”brokeh fitting in the weigh cell tubing was ;repaired. The relationship between the weigh cell readings and the loca- :rtions of the probe 1ights was estdblished. The weighing system on. the coolant drain tank proved to be con31derably | V;fmore steble than on the fuel drain tanks. With 5756 1b of salt in the tank at ebout 1200°F the extreme spread of 40 1nd1cated welghts, taken over a period of a week, was t 22 1b. e o A e " S 160 6.8.4 Heatup and Cooldown Rates These are described under heaters in Section 17. 6.8.5 Discussion snd Recommendations “ | : N6 unfisual difficulties were encountered with the coolant drain tank. The weigh cells fUnctiohed satisfactorily, however, the trouble with the fuel drain tank weigh system emphasizes the need for cereful consideration of piping stresses or the use of other type instruments. 161 7. COVER-GAS SYSTEM P, H. Harley | 7 The MSRE cover-gas system consisted of a helium trailer for normal sup- ply, two banks of three helium'cylipders.forremergency_use, two parallel ,treating_stations_for_rempvingrmoisture and oxygen to a nominal 1 ppm by 7_roiume, a treated helium surge tank, and two hesders,_35 psig,end-250-psig, ~which supplied cover gas to the various. systems. Each of the treating sta- tions consisted of a dryer (molecular sieve) preheater, and an oxygen- removal unit - (heated tltanium sponge) Mbisture and oxygen anaiyzers could be valved in to monitorrtheetreated or untreated helium as desired. 7.1 Initiel Testing The components of the treating station were developed and designed by the Reactor Division Development Section. The'performance of the sys- tem was tested by measuring oxygen and moisture content of the cover gas. upstream and downstream;of:theetreating station to insure proper opera- . tion. Other testing at the MSRE consisted of 1eak-testing the system and , checking flow control and temperature controls, Thermal expansion of the 1 0, removal ‘beds during the initial heatup caused the 3-in., ring joint flanged heads to leak, After retightening the flanges, the units were kept hot to prevent thermal cycling the flanges.f , | i Late tests indlcated that the Fisher Mbdel 67—H pressure regulator }.which was installed to reduce helium from traller pressure to 250 psig was _}susceptible to moisture diffu51on through the diaphram., These tests indi- ) ';_cated Victor.Model VPS-201 - and Grove Model RBX-20h-15 were better from . 162 this standpoint. The Victor model was less complicated and therefore was installed 1n the MSRE. The low-pressure cover-gas header was 1n1t1ally a hO-p31 header which was protected by a U48-psig rupture disc. During initial testlng, this rup- ture disc failed on several occasions withoutVapparent'cause; The header pressure was subsequently lowered to 35 psig'which was sufficient for all normal operating requirements and did alleviate the problem, The safety 1imits2% (Lo He supply pressure) had to be lowered from 30 to 28 p51g to provide s margin from the normal operating pressure. Four helium control valves failed in the first month of operatlon. Some galling of the close flttlng trim (17-4 PH plug and Stelllte seat) was apparently caused by the extremely clean, non—lubrlcatlng_dry helium. Other valves in the system, inclu&ing the four replacements have operated satisfactorily. 7.2 Normal Operation Since putting the cover-gas system in operation in the fall of 196k, sixteen trailers of helium have been used. About 21,800 ft3 of the 29,000 ft3 in each trailer was used before returning the trailer to be re- - filled. A trailer lasted about 4 months on the average or an average usage of approximately 180 cubic feet per day. Normal usage when the reactor was - operating was 260 cubic feet per day. Larger amounts were used during pur- ging and filling procedures and very little was used during maintenance periods. Although helium was the normal éover'gas, during maintenance periods:" ‘fhe reactor system was purged-with argon. Argon was also used during a special study of circulating gas in the fuel. Bottles of argon containing 5 psi during Run 4) was negligible. Exposure to atmosphere could have opened some of the pores in the filter; however, a more likely - explanation is that handling and sampling opened enough pores in the ele- ment to produce these results. 181 Fig. 8.6B PCV-522 Valve PHOTO R27812 182 8.4.2 Run 5 o Satisfactory manuél‘presshre—control by throttling V-522B was demon- strated during theipreceding shufidown; however, at the beginning of Run 5 after only 2-1/2 hburs of power bperation (1 MW), this valve showed evi- dence of rapid plugging and required comstant adjfistment until it was in the fully-open position fifie hours later. During one such adjustment,' rapid plugging of the charcoal beds occurred. At the conclusion of salt circulation (terminated by a space-cooler motor failure), the restriction at the inlet to charcoal bed 1B could not be dlearéd; therefore, the inlet valve (HV-621) was removed for examination at the HRLEL. With the valve out, the pressure drop aérossrthe bed was much lower but it was still higher than for a normal unrestricted bed. An excess of heiium (forward blow) was forced through the bed and the pressure drop suddenly decreased to normal. The cone of the inlet valve to this bed (shown in Fig.78.7) was shiny as though wet with an oil-like material. The small metallic chips near the large end of the cone were a result of the sawing operation. There was some white amorphous powder on the tapered section o0f the valve trim and appeared to be adhered fairly strongly to the metal surface. A similar material was found in the valve body. The material removed from the stem was described as an isotrbpic,.faintly colored material, varnishlike in ‘appearance with a pebbly surface. The predominant isotope in the material was ***Te. The refractive index of the varnishlike material was 1.526 com- pared with 1.50 for a distilled fraction of the lubricating oil used in the fuel pump.* ' | 8.4.3 Experiments and Alterations During the March-~1966 Shutdown Off-gas samples taken while the reactor was shut down showed an in— creasing hydrocarbon content ifi]the gas as the reactor cell temperature was - increased, lending support to the'hypothesis that there was a reservoir of hydrocarbons in the holdup—voluhe portion of the off-gas line. It was not practicable to clean the‘68-ft-1ong, 4-in.-diam pipe, so the off-gas line was disconnected at the fuel pump and in the vent house; large quantities of helium were blown through the line in the forward and reverse directions at velocities up to 20 times normal.®»” Very little visible material was 183 ~ PHOTO R28500 184 collected on filters at the ends of the line, but there were fission prod- &fi# ucts, and the amount doubled when the cell was heated from 50 to 80°C. Visual observation showed that the head end of the holdup volume was clean except for a barely perceptible dustlike film. A thermocouple was attached to the holdup pipe near the head end for monitoring temperatures during power operation. (When the power was subsequently raised, the temperature 'indeed rose from cell air temperature of about 55°C at zero power to about 113°C at 7.5_MW; The temperature increase occurred with a time constant of about 30 min which was not inconsistent with buildup of gaéeous fission products in the line.) | | | - As a result of extensive laboratory tests on orgénic vapor traps,’ the filter-valve assembly was replaced with a particle trap and an organic vapor trap in series and upstream of the pressure control valve (PCV-522) | whose flow coefficient was increased to 0.865. The Mark I pérticle trap,’ shown in Fig. 8.8 was designed to remove particulates and mist. .Gas from the fuel-pump bowl entered at the bottom of the unit through a central pipe, reversed flow in the Yorkmesh, and passed in succession through two con- o centric cylinders of porous metal (felt metal) and a bed of inorganic fibers. kfig The first felt metal filter was capable of stopping 96.7Z of particles | greater than 0.8y and the second — 99.47% particles greater than 0.3p. The organic vapor trap’ consisted of713 feet of 1-in. sch.-40 stainless- steel pipe, arranged in three hairpin sections of approximately equal lengths. The vapor trap was loaded with 1092 gm of PittsBurgh PCB charcoal and instru— mented with thermocouples at 5, 12, 51, 59, 105, and 113 inches, respectively, from the bed inlet. | As a result of the experience with and examination of HV-621 and other off-gas components, the inlet valves to the main and auxiliary charcoal beds and V-522A were replaced with ones having larger flow coefficients. 8.4.4 Runs 6 and 7 At the start of Run 6, the pressure drops (at normal flow rate) across the particle trap, organic vapor trap, and charcoal beds (sections 1A and 1B in parallei service) was <0,05, 0.7, and 1.6 psi respectively. The off- | gas system performance was satisfactory during.the first ten days of Run 6 with the reactor operating at 1 M{ or less. However, when the power was Y - FINE METAI‘.L'IC_ FILTER‘ FIBERFRAX ORNL-DWG 66-11444R COARSE METALLIC FILTER'; E (LONG FIBER)N © -+ . —THERMOCOUPLE (TYP) ’ s A = """"5 3 7z '“'/ flw,s /d’/.@'/fl/////fi'/mflfl //l/l/;/ ¥ ||5|'<|a?3 N - 9 i i 2 A m | ‘ ‘ l ' il i fi E T ey m;: wum e, mr::,;:::_::a : QIIIIIIhLI I b ! rgwg ,gfi ' 4 b L e ’*r*I?’ifilm 7 g Ao ¥ % rxrat /‘9" { 7 b 1 z Zz2 u”u””u/uu//fl' //fl”.wf// ////// ////4’///#/ /A”//////”./ll// - ‘ : S oM : o it ~~SAMPLES A LSAMPLE 3A L8 5B AND 6 | STAINLESS STEEL MESH NICKEL BAFFLE (8) Fig. 8.8 Line 522 Particle Trap (Mark I) and Location of Sampling o Points During Subsequent Examination 68T 186 increased to 2.5 MW, the pressure drop across the charcoal beds also in- Qfij creased. Pressurization and equalization experiments established that the restrictions were at the inlets to'the’beds,.probably at the packing of » steel wool above the charcozl.- It,was‘found that the pressfire drop could be reduced, usually to near the normal 1.5 psi, by blowing helium at 35 psig either forward or backward through the bed; back-blowing seemed to be more effective. Back-blowing of the beds was done whenever the pressure drop of two sections in parallel'approéched.B psi. Later, higher pressure drops were tolerated before baCk-blowing was initiated. Plugging occurred when thé power was raised during the épproach'to full power with section 1B plugging more often than the others. Plugging became less frequent later, but at the end of Run 6, it was still necéssary to back-blow the beds about once'a week. Particle Trap Performance. After approximately 10 days of operation at the 5-MW level, the pressufe'drop across the particle trap started to increase and reached 6 psi abprbximately a week later. When the powet was increased to 7.5 MW, the pump pressure increased to 9 psi in less than 12 hours at full power. Since the pressure drop across the organic vapor trap (fiJ was less than 0.5 psi and that across the charcoal bed was kept below 3 psi by periodic back-blowing, the variation in pump-bowl pressure was inter- preted as a cdrresponding restriction in the particle trap. About three hours after an unscheduled power interruption, the fuel-pump pressure de- creased to 6.5 psig and the following day it decreased to 4.3 psig. How- ever, a few hours after the reactor power was again raised to 7.5 MW, the pump pressure increased to 8 psig indicating a restriction in the particle trap. At this time the helium purge to the punp shaft was decreased from 2.4 to 1.9 liters per minute in an effort to reduce the fuel-system pres— sure. The reactor was then taken subcritical for two days. When the power was again increased to 5 MW, the pump pressure increased from 3'psig at zero power to 8 psig and eventually to 12 psig at full power. The pump pressure gradually returned to the 6 to 8 psig range after approximately a week at full power. After Run 6 was terminated by a componenf cooling pump failure on May 28, 1966, the particle trap pressure drop decreased to - less than 1 psi and it remained relatively unrestricted for a month after full-power operation was restored in Run 7. The pump pressure then gradu- Qfiyf ally increased to 8.7 psig before another power reduction (MB-1 failure) 187 occurred'ending Run 7. A few hours after the power reduction, the pump pressure returned to a normal value of 3 psig and remained at this pressure - for approximately a month of zero-power operation before the restriction ~ gradually returned. ' . In summary, pump pressure increases (restrictions in the particle traps) were usually associated with periods of power increases. Pressure decreases were sometimes unexplained but usually were associated with power .- .decreases or'with'deliberatelattempts-to clear the trap such as reverse- blowing with helium. - ConEny ) | | Rerouting of Line 524 Helium, at a rate of 2.4 %/min, was introduced through line 516 into the'fuel—pump shaft annulus just below the lower shaft oil seal;a:Thejlargerjpart-of.this flow'goes'down the shaft into the pump bowl thuszpreventing fission gases and salt mist from entering this radi- ‘ation sensitive region of,thé"fiuap. 'The smaller portion of this flow (<0.1 2/min) goes up the shaft to prevent oil vapors from migrating to the 'fueiusait;'it also aids in”transpOrting any oil seal leakage to the oil r_catch tank. The driving force;for the smaller flow (line 524) is the pres— sure difference between the fuel pump and line 522 downstream of the fuel - .system pressure control valve (PCV-522). However, when PCV-522 was opened fully (because of restrictionS'in'the charcoal beds), line 524 had no pres- sure differential and consequently no gas flow; therefore, it was rerouted - to enter the off-gas line at V-560A downstreau of the main charcoal bed and V-557B. V-557B was thenrthrottled to control system pressure. On two oc- | casions in Run 6, sudden increases in pump pressure (at the completion of salt recovery from the overflow tank) caused a gas flow reversal at the pump- shaft and gaseous. activity was carried into line 524. Since this line -bypassed the charcoal bed the gas flow had very little decay time. As a *fj;consequence, gaseous- activity triggered -the . closure of the radiation block -~ valve on the main off—gas line. Therefore in June of 1966 'a small charcoal 1bbed’ was added to 1line 524 to: hold “up krypton and xenon for 2-1/2 and 30 days, respectively.r The bed - consisted of 9 ft of 3—in. sched -10 stainless—: - steel pipe loaded with 15.8 lb of Pittsburgh PCB charcoal On at. least four : occasions in the subsequent run, fission product activity was blown or dif- -'fused up the»pump shaftrannuius.i Although the activity level in the oil 'catch'tank-and line 524 increased, there was essentially no activity re- iease, thus proving the effectiveness of the added charcoal bed. However, 188 the last two releases from the pump bowl caused the flow element in line o 524 to plug partially and then completely. The plug was fbund to be in ‘the sintered stainless-steel disc at the inlet to the matrix—type flow ele- ment. The element was replaced with a capillary tube. 8.4.5 Expgriments and Alteration During the June-September Shutdown (1966) At the conclusion of Run 7, it was decided to flush out ;he'residual fuel in the overflow tank and to check the indicated level at which over- fldw occurred. Because of an insufficiency of salt in the drain tank, pres- surizing gas from the drain tank entered the reactor and floodéd'the pump bowl with salt, thus binding the pump shaft with frozen salt and pushing salt in the gas lines at the top of the pump bowl. | - The frozen salt was cleared from the reference bubbler line with the use of femotely—applied external heaters; salt in the sampler line was melted by the same technique but it ran down and refroze at the junction with the pump bowl. This obstruction and the frozen salt in the énnulfis around the pump shaft were cleared when the pump bowl was refilled and heated to 650°C. Although some salt entered the main off-gas line as indi- cated by a temporary rise in temperature at TE-522-2 (Fig. 8.9), pressure ‘E,) drop measurements showed no significant difference from the clean condition. This was attributed to the blast of compressed helium from the drain tank that was released backward through the off-gas line, before the salt had time to freeze completely, fihen the overfill triggered an automatic drain.® Therefore the only work done on the main off-gas line (522) during this shutdown was to replace the short, flexible "jumper" section of the line, where the convolutions would be expected to hold some salt. The particle trap was also removed from line 522 and sent to HRLEL for detailed examination. An identical replacement (Mark I) was installed to permit operation while the original was examined and a new one designed and constructed. Also during this shutdown, a remotely replaceable heater was designed and installed on the inlet section of the auxiliary charcoal bed (ACB). 8.4.6 Runs 8, 9, and 10 With the resumption of operations the off-gas flowed freely, with no unusual pressure drop for 26 days of circulating helium,‘flush_éalt; and 189 - ORNL-DWG 67 - 4764 OFFGAS HOLDUP VOLUME , 4in. DIAM TO PARTICLE TRAP - J AND CHARCOAL BEDS ==—}h | / Fig. 8.9 Off-Gas Piping Near Fuel Pump and Overflow Tank 190 fuel salt at low power. Then, tfio days after power operation was resumed at 5.8 MW, a plug developed in line 522 somewhere between the pump bowl and the junction of the overflow tank vent with the 4-in. holdup line. The first indication was a decrease over a few hours from 107°C to 71°C at TE- 522-2, as the plug caused the off-gas to bypass through the overflow tank (OFT). The presence of the plug was confirfied when HCV-523 was closed to build up pressure in the OFT to return salt from the tank to the pump bowl; pressure in the pump bowl also built up. Efforts to remove the restriction by applying a 10-psi differential either forward or backward were unsuc- cessful. ) | The bypassiné of'off;gasyihfofigh the OFT did not hinder operations except for one specific job: recovery of salt from the overflow tank. Salt sidwly but continuously accumulated in the tank during the entire operating life of the MSRE, and it was therefore essentiél to return salt to the pump bowl two or three times a wéek in order to maintain proper levels. With a plug in the off-gas line of the pump bowl, it was necessary to greatly reduce helium flows into the pump so that the overflow tank pressure could be increased faster than that in the pump bowl to make the salt transfer. Through the remainder of Run 8, salt was returned (burped) from the OFT six times, and on at least four of these occasions some fission product activity was blown or diffused up the pump shaft annulus into the oil collection sp;qe. This was a consequence of the reduced helium purge down the shaft annulus and the uhavoidable,sudden preésurization of the pump bowl that occurred at times in the procedure. After Run 8 was terminated, steps were-taken to clear the plug from the off~gas line so- that the normal salt recovery proceduré could be used. Frozen salt was suspected.as the cause of the plug, so the fuel loop was flushed to reduce radiation levels, the reactor cell was opened, and speci- ally built electrical heaters were applied to the line between the pump bowl and the first flange. Heating alone did not clear the plug, but when, with the line hot, 10 psi was applied backward across the plug, it blew through. The pfessure drop came down as more helium was blown through un- til it became indistinguishable from the normal drop in a clean pipe. In Run 9, the power operation was begun 8 hr after fuel circulation had commenced, but TE~522-~2 came up to only 66°C, indicating that the line 191 . was again plugged. Whileftools"and procedures were being devised, the re~- ~actor was kept in operation, but great care was taken to avoid getting fis- “sion products or salt-spray7up the pump shaft annulus again., This entailed ',lowering ‘the power to 10 kW, 24 hr before the overflow tank was to be emp~- tied, then stopping the pump 4 hr beforehand to let the salt mist settle. During the salt transfer, the fuel pump was vented through the sampler and =auxiliary charcoal bed. After three cycles of this, the reactor was drained and flushed again:in preparation ‘for working on the off—gas line. This time heat was applied to the short section of line between the second flange and the top of the 4—1n. decay pipe. 'When heating to about 600°C did not open:the line, the flexible jumper was disconnected to permit “clearing the obstruction mechanically. In the flange above the 4-in. line, the 1/2~-in. bore was completel& blocked, but the weight of'a:chisel tool broke through what appeared to be only a thin crust of ‘salt. 'Borescope inspection showed that the rest of the vertical line was practically clean,r and there was only a thin layeriof salt in the bottom of the horizontal rum of the 4-in. pipe. Helium-was:blown through the line at five times the normal flow, and the'pressureidrop indicated no restriction. 'The flange ‘near the pump bowl contained;arsimilar plug which was easily broken. A 1/4~in. flexible tool was . then inserted all the way into the pump bowl to prove that a good-sized passage existed A new jumper line was installed and operation was resumed., R o | ‘Because obtaining samples remotely without spreading contamination - would have been most difficult, no analyses were made of the material in . the flanges.- But it appeared that salt: had frozen in the line almost com- pletely blocking it during the overfill of July 1966. Material in the | ';foff—gas stream during subsequent operation ‘then plugged the small passages. ‘During the July cleanout the heaters apparently melted the salt out of the 'frpipe, but left behind a thin bridge of salt in the cooler flanges. After re'the flanges were mechanically cleaned and reassembled “this’ part of the -doffngas system operated satisfactorily during Run 10. 8. 4 7 MK-I Particle Trap 1~?l**€i‘f' ‘ ) : The second particle trap -served through Runs - 8 9,_and 10., This'unit " behaved in Runs 8 and 9 much as had’ the first trap; the pressure drop oc- - casionally built up to 5 to 10 psi, beginning two ‘days after power operations 192 started in Run 8. Back-blowing with helium-was effective in reducing the pressure drop to 2 to 4 psi; however, in Run 10, after the first week of power operation, back-blowing became ineffective. Various tacticsa were used to get fission gas to the particle trap with as little delay as possi- ble, to see if increasing the fission product;héatifig in the trap would drive off the material at the restriction. After this proved to be inef- fiegtive and recognizing that heating caused the central inlet tube to ex- pand farther into the Yorkmesh filter in the trap, the opposite- approach was used. Eight hours of delay was obtained by routing the gas through an equaLizing line to the empty drain tank, through the tank and salt fill 1lines to the other tank where it bubbled through several inches of salt heel, then out through the drain-tank vent line to the particle trap. The pressure drop across the trap was 16 psi when the gas was first rerouted, but within a fewfihours it was below 2 psi. When the original'route was - again ;ried,.the restriction began to build up almost immediately; there- fore,.;he delayed route was used until the end of Run 10 which was termi- nated for in-cell repairs and also because a newly constructed particle trap (Mark II) based on the Hot Cell examination of the first trap was - ready for installation. | ) . S Examination of MR-I Particle Trap.® Flow tests on the first trap (in ser?icg from April through July of 1966) indicated that (1) the pressure. drop after service was about 20 times the "clean" pressure drop. .Assuming' orifice type flow, this would represent a reduction in flow area of about 80%. (2) The bulk of the pressure -drop occurred in the Yorkmesh and Felt- me;al_sectioqs,of the filter. The Fiberfrax section was essentially clean. The area of the Yorkmesh which had been immediately below the inlet pipe was cqvered with‘a blue~gray to black mat which hadrcompletely filled tfie.spéce between the wires of the mesh (Fig. 8.10). The shapeaof the mat éorresponded to the bottom of the inlet pipe, and it is likely that this matérial was the major restriction to gas flow while in service. Since the inlet pipe temperature probably increased several hundred - degrees during power operation of the reactor, it is believed that this restriction behaved similar to a thermal valve. This could account for the unexpected increase in pressure drop while at power and the decrease in . pressure drop when the power was reduced or when the off—gas‘was_deléyed | via the drain tank. Fig. 8.10 193 : Ttap Mark I 194 A radiation survey made around the outside surface of the lower section of the trap gave readings ranging from 6900 to 8000 R/hr. The probe indi- cated 11;400 R/hr when inserted into the position formerly océupied by the inlet pipe. The radiation level dropped off to 4100 R/hr at the bottom of the trap. The bottom of the inlet pipe had a deposit of blackish material which corresponded to that in the Yorkmesh. The inside of the inlet pipe at the upper end of the bellows appeared clean and free of deposit. The eXtétior of the bellows had some of the light-yellow powdery material on a background of dark brown. = ' When the Yorkmesh was remofied, it was found that the surface of the outer wires of the mesh bundle was covered with a thin layer of amber- colored organic.materiél.' Much of fihis material evéporated from the heat of the floodlamp used to makéfthe photographs. As the bundle was unrolled, the color of the film on the fiire changed from amber to brown to black near the center. The black material waé thicker than the wire by a factor of 2 ‘or 3. This material was brittle, as was the wire, and much of it came -léose as the wire was flexed. Samples of this material,’designated Nos. SB and 6 (Fig. 8.8) were taken for examination and chemical analysis. Met- éllographic examination of a piece of the wire covered with the black ma- terial showed that the wire was heavily carburizgd with a continuous net- work of carbide in the.grain boundéries; There was no evidence of melting of the wire; however, the grain growth and other changes indicated oper- ating temperatures of at least 650°C. The nonmetallic deposit observed on the wire mesh was apparently of a carbonaceous nature and appeared to have been deposited in layers. These "growth rings'" were probably the result of off-gas temperature, reactor power, and gas flow-rate changes. The perforated plate of the coarse filter section and the lower flange of the filter assembly were covered with a stratified scale (view A-A, Fig. 8.8). The colors varied from a very light yellow to orange. One stratum in the lower flange area appeared gray, almost black. A sample (No. 9) was taken of the light-colored material, including some of the black material. The perforated plate of the fine filter section was covered with a8 thin, dark-brown coating, which seemed to be evenly distributed over the surface of the plate. The inner surface of the outer wall of the trap was covered with allight—amber coating, which was also evenly distributed. - s | 195 It is believed that_these coatings were deposited by condensation and sput- :tering.of the oil from the adjacent filter and that the dark brown color indicates that the porous'metal screen had operated at a'much higher tem— .perature than had the- outertwall (during operation the trap was immersed 'jin a tank of water for cooling by natural convection) ‘The lower surface 'of the upper flange (view BwB Fig. 8.8) contained a depoeit Which'had the -tappearance of organic residue.s,The deposit was amber colored, and the frac- 'ftured edge (Fig. 8. 11) gave the impression that the material was brittle. ‘There is as yet no explanation for formation of this deposit or how it came to be formed in this particular location, The radiation level on the out- _'side of this section of the filter was 200 to 360 R/hr. '_ The material at the entrance of the Fiberfrax section: showed an oil- '- likefidiscoloration, but there was no. evidence of any significant accumula- Vtion of ‘material. Comparison of the weights of the different layers with ‘the weights of the material originally loaded indicated changes of less than 0.2 8 An interesting observation relates to the very ‘low radioac- tivity level of the Fiberfrax at the entrance section which is separated "only‘by the Feltmetal filter from an area containing material with activity - levels: of thousands of R/hr._ The only detectable activity (above the exami- ination cell background of 4.2 R/hr) was, at the discharge end of the Fiber- frax section. It is probable that this activity resulted from back—blowing the trap or from pressure transients which could have carried gaseous decay 'products from the vapor trap back upstream and into the particle trap. | .;Even so, the activity level was only 1.8 R/hr above background.__ 7 Analytical Results. | A total of four samples from three different lo- ' cations (Fig. 8.8) were subjected to a variety of analytical tests. The samples were identified as follows. - _S ple No.-i'j_ ’jEf' ], o " Taken from . | 3A ..;_lf-r'-';Coarse section of porous metal filter 9 . Scale on lower flange of porous metal section 5B % Mat at inlet to Yorkmesh section 6 ';Li'Mat at inlet to Yorkmesh section - . For sections of sample 3A it was found that about the same weight loss (0.2% of sample weight) resulted from heating to 600°C in helium as from _ dipping in a trichlorethylene bath. The material removed by heating was 196 11 Upper End of Porous Metal Section — MSRE Particle Trap Mark I 8 Fig. . 197 Rl s cold trapped and found to be effectively decontaminated; however, the tri- chlorethylene wash was contaminated with fission products. | Samples 5B and 9 were compared for low-temperature volatiles; at 150°C, No. 9 lost 52 and No. 5B lost none. When raised to 600°C, the weight losses were 35% for sample 5B and 32% for sample 9. "Analysis of the carbon content gave none for sample 9 and 97 for sample 58. This indicates that sample 9 'had not reached as high a temperature as had sample 5B. ' The mass spectrographic;analysis»of sample 6 indicated that there was a very high fraction'of'fission products. These are estimated to be 20 - wt % Ba, 15 wt % Sr, and 0.2 wt Z Y. In the same analysis the salt con- stituents Be and Zr were estimated to be 0.0l and 0.05 wt % respectively. In addition~the'materiel-in'samples 5B and 6 contained small quanti- ties of Cr, Fe, and Ni, while sample 9 did not. The reliability of these values was compromised by difficulties caused by the presence of organics and small sections of wire in the sample. ‘The gamma-ray spectrographic work indicated therpreSence of the fol- lowing isotopes: '®7Cs, ®°Sr, *°°Ru or *°°Ru, '*°MAg, °°Nb, and *“°La. All rhree samples were chemically analyzed for Be, and the level was below the detectable limit of 0.1l%. Attempts to analyze for Zr were com- plicated by the presence of large quantities of Sr. ' The fraction of soluble hydrocarbons was determined using CS:, and the values were 5B, 60%; 6, 73%Z; and 9, 80%. The extract solutions from ssmples 5B and 6 were allowed to evaporate, and a few milligrams of the residue was mounted between salt crystals for infrared analysis. The sam-— ' ples were identical and were characteristic of long-chain hydrocarbons. There was no evidence of: any functional groups other than those involving carbon and hydrogen. Nor was there any evidence suggesting double or triple bonds. There was an indication of a possible mild cross-linkage. It is likely that there is more cross—linkage of the organic in the gas stream' than appeared in these samples, and the low indication could be due to the {{insolubility of the cross—linked organic and the high operating temperatures of the wire mesh, which would,cause breakdown of the organic into elemental carbon and volatiles. 198 8.4.8 MK-ITI Particle Trap As a result of the operating experience with and detailed examination of the first particle trap, the MK-II trap,® (shown in Fig. 8.12) was de- signed with the following features: - 1. The trap housing was increased from 4 to 6-in. ID, resulting in an increase in cross-sectional area of 2254 in both the Yorkmesh and Fiber- ~ frax sections, . , 2. The unit was in effect turned upside down so that the Yorkmesh section is at the top of the unit and the Fiberfrax section is at the bot- tom. This change permits heating of the Yorkmesh section (using beta decay heat -and lowering the water level) while still maintaining cooling on the . other two sections. ' 3. The disposition of the Yorkmesh was modified to provide increased frontal area in.the directien normal- to the flow. 4. Since the first trap had shown veryrlittlerloading-infthe,fine Feltmetal section, only the coarse Feltmetal was used. , 5. The total filter area was increased from 22 in? for the original to 288 in® for the new trap. 6. The depth of the Fiberfrzx was reduced by 50%. 7. The pressure drop at 15 2/m was less than 1 in. Hz0 and the trap . < efficiency was 99.9Z for particles greater than 0.8u. 8.4.9 Organic Vapor Trap It was expected that accumulation of orgenic material in the charcoal trap immediately downstream of the particle trap would result in progressive poisoning along the length of the trap. Such poisoning would shift the lo- ‘cation of maximum fission product adsorption and produce a shift in the temperature profile of the trap. Except for the upward shift due to in- creased power level, the basic shape of the temperature profile did not change, indicating that no significant poisoning by organics occurred. Since the pressure control valve (PCV-522) was operated in thelfully— )opened position and it appeared that nothing would be lost by eliminating the charcoal trap, both were removed to meke room for two new MK-II parti- cle traps which were installed in parallel in January 1967. ORNL~DWG 67-4766 1 0 i 2 3 INCHES _ YORKMESH A | L . /_-T.E. _ /FELTMETAL . N /T.E. . \FIBERFRAX . 'FE. L Sl 2 AL L L [L T PP P T f e L — [~ re—— ! Vo T A e = = et - — - ’ ———— et B L \\ N Y s/ - ' ’%’ 7 RS o ~ ™~ A 2 t rF VY ! —: —— —i e A. & y 4 ) T = / 7T 77 777 77 777 7 L 1] : - NICKEL BAFFLES s, Fig. 8.12 Line 522 Particle Trap Mark II, Particle Trap Subéssembly 66T e g e g g et pump bowl pressure) to force the salt back to the pump bowl. “When the OFT discharge is opened, the salt remaining in the tank and that remaining in the pipe is Sufficient to provide a seal.at.the bottom of the tank and thus prevent-fission_product;gases from'entering the overflow tank. However, when the main_off-gas line, 522, became restricted near the pump 'bowl; practically all of the pump purge gas and gaseous fission products were bubbled through the salt heel in the OFT and out through line 523 to join the main off-gas line downstream of TE—522 2. From TE-522-2 data and pressure measurements during salt recovery operations, it was estimated that fission.product gases were first diverted through the OFT on.October 15, 1966. - Since the fuel—pump overfill on July 24, 1966, it is estimated that ' ‘most - of the fuel—pump gases were forced through the overflow tank for ap- proximately 28 days in 1966 6 days in 1967, 76 days in 1968 and 68 days in 1969, before the overflow tank gas exit line (523) became plugged on -rMay 25. Thereafter, all of the,pump purge -gas was forced.out through the . restriction inlthe off-gasilinerat-the exit of the.punp bowl. This situ- ation caused some inconvenience duringrthe salt recovery operation,'but by oreducing-purge flow to the:pump atothese times, the operation could be done -without exceeding 15 psig in- the pump bowl. | | | During the shutdown in June 1969 when the overflow tank vent line was scanned with the remote: gammaflspectrometer,_an.unusually strong source was Tobserved at the air;operatedrvalve:(HCV;523), about 33'ftudonnstream from ;-the'overflow tank. ?langes'fiere'opened~and pressure observations showed that the restriction was in the flanged section containing both the air- 'operated valve and a hand valve. This section was removed to a hot cell where polymerized hydrocarbons and fission products were found to be block- h'ing the hand valve inlet port. A replacement section containing,an,air— ‘operated valve but no hand:valve;was:installed;_ Although the restriction : atrthe exit of the pump bowl reappeared during the last week of MSRE opera- tion, the OFT discharge line functioned normally. 210 8.5.7 Restrictions at Exits of Drain Tanks During the reactor drain in June of 1969, pressure measurements indi— cated that a restriction had developed in the gas line somewhere between Fuel Drain Tank No. 2 (FD-2) and the junction of the inlet line (574) and the exit line (575). The first half of the drain appeared normal at a drain rate of v2 cu ft of salt per minute after which it tapered off to m0,25 cu ft per minute when the trapped gas in the drain tank balanced the salt in the fuel system. Two and a'half hours were'required for the drain as op- posed to a normal drain time of less than 40 minutes when drained into one tank. The restriction was later cleared by heating the tank to 680°C and applying a pressure‘differential of 60 psig across the plug via line 561. A similar but lesser restriction in'the’gaszline at the.other fuel drain ‘tank was partially cleared during the fuel system pressure'test in'Auéust 1969 by heating the tank to 680°C and flowing helium from FD-2 (at 50 psig) to FD-1 (at 2 psig). , Restrictions in these lines were most probably deposited (but not to the poiut of detection) durihg December of 1966 and January of 1967 when the reactor off-gas was routed to FD-1, through FV-106 and FVAIOS,Vthrough the salt heel in FD-2, and out through line 575 (Sect. 5.6.1). Fission gas decay heating, during this time, was appreciable in the drain tanks and re- quired downward adjustments to the drain tank heaters. The offegas £low through the drain tanks was reversed after the first week in an effort to minimize plugging. - . | | , The off-gas flow was again routed through drain tank No. 2 gas lines for a day when line 522 plugged in May.of 1969. | In October of 1969 during a pressure release test When the fuel—pump 3 gas was released into FD-1, it was discovered that the gas line was again restricted near the drain tank. Two attempts were made to clear this line by back-blowing helium, first at 30 psig and then at 50 psig, from line 561 through HCV-573 and into FD-1. Although only marginallimprovement was ob- ~ tained, the gas flow through this line was judged adequate for any emer- gency drain situation. ‘During the subsequent drain in November 1969, approximately 4300 1b of salt drained into FD-1 and 5100 lbs drained into FD-2 iudicating that a slight restriction existed in the exit gasvline of FD-1. During the . 211 December -drain, the‘exit,gas valve from FD-2 was closed during part of the. drain to compensate forrthe partial restriction in FD-1 gas line; however, overcompensation resulted and 5400 1bs drained into FD-1 and 4100 1bs drained into FD-2. The drain times for both drains were less than 20 min- utes which is normal for a drain when both tanks are used. 8.5.8 Restrictions at,therbffégas Sampler 7 A system to permit the analysis of the reactor offegas stream was in- stalled downstream of the particle trap as shown in Fig. 8.2. The sampler contained two thermaleconductivitycells, a copper oxide converter, and two molecular Sieves,'one operating atfiliquid nitrogen temperature and the other at room temperature.® Since the sampler wasran integral part of primary containment during samplingboperations, and since some components of the | sampler did not meet the requirements of primary containment solenoid - block valves were installed in the inlet and outlet lines which connect - the sampler to the reactor .system. ‘Two fail—closed valves in series were -installed in each line and were instrumented to close on high sampler ac—~ | tivity, high.reactor'cell-pressure, and high fuel-pump pressure. | , Although located downstream'of the particle trap, the inlet line to the off-gas sampler periodically developed a restrictionrin the vicinity of'the'safety block. valves. The block valves have 3/32—in.-diam ports and the inlet piping is 0. 083 in ID autoclave tubing. The restriction was suc- cessfully cleared each time by back—blowing with helium, Aslencountered in other parts of the off—gas system, successively higher pressures were _required to clear the restriction each time. During the latter part of Run 18, the inlet block valves would not shut of f tightly and were replaced in July of 1969. Visual. inspection of the faulty inlet valves did not re- 'veal the .reason for the. leaking valves nor- the nature of ‘the restriction "(the restriction had been blown: clear before the valves were -replaced). _During the August startup, 1t was again necessary to blow out the restric— tion at the inlet to the sampler.- 212 8.6 Discussion and Conclusions The periodic plugging in the coolant system off-gas filter and also in the fuel system filter during the precritical and low-power (<25 kW) operation can be attributed primarily to the accumulation of oil in the 0.7 to 1lu diameter pores of the sintered metal filters (2 to 4p pores might have been a better choice). However, subsequent filters inrthe fuel system were constructed of felt metal (Huyck Nos. FM-225 and FM-204) which was not only efficient at stopping solid fission products but also was apparently immune to plugging by hydrocarbons. Virtually all of the solid decay daugh- ‘ters were stopped by the Yorkmesh and filter before reaching the Fiberfrax section of the particle trap. 0il vapor in the off-gas stream condensed on various parts of the off- gas system components and apparently enhanced the adsorption or trapping of particulate matter, particularly on the Yorkmesh portion of the particle trap. | | 7 The fact that such large amounts of solid decay daughters were trapped in the particle trap indicates that most of these solids tend to remain in the gas phase until trapped onto a surface wetted by hydrocarbons.f It is not clear whether the solid decay daughters agglomerate in the hold-up vol- ume and if so whether they agglomerate with others of their own species, with other species, with hydrocarbons, or‘possibly combine chemically to form both volatile and non-volatile compounds.r A large fraction of the noble gas fission products decay in the 4-inch-diameter hold-up volume and their decay daughters apparently plate out or are adsorbed onto the first flow channel restriction where the gas flow changes'direction"andlor is close to a relatively cold surface such as valves and floufrestrictors; | The high hydrocarbon content (>60%Z) of the material collected on the Yorkmeshris inconsistent*with.the high temperature (¢650°C) in this region | during power operation; hydrocarbons are vaporized and cracked at this tem- .perature. One must assume, then, that the hydrocarbons were collected sometime after power operation had ceased and after the fission product de- cay heat became negligible. The growth rings on the Yorkmesh were probably caused by alternate periods of power and zero-power operation. 213 Since spectrographic analysis did not show any of the alkali metals, cesium or rubidium, on the Yorkmesh, they were apparently boiled off also by fission product decay heating in this region. If this be the case, ap- preciable quantities of these materials would be expected to be collected in the charcoal beds, On the other hand,; large quantities of barium and strontium were reported'on'the Yorkmesh; this would argue for effective trapping of their precursors, cesium and ribidium. However, to be trapped on Yorkmesfi.at 650°C, thelalkali:metals would necessarily have to be chemi- cally combined (as halides for instance) rather than in the elemental form, otherwise they would boil off and be carried downstream. _ Although none of the solid decay daughters in the 4-in.-diam pipe got past the particle trap, another batch (though somewhat smaller number) of solid decay daughters are born (and presumably remain in the gas phase) in the larger hold-up volume ahead of. the main charcoal beds. These fission products along with hydrocetbdné-are;probably the source of the plugging experienced at the entrance'region‘of—the beds. After examination of the MK-I particle trap, it is understandable why the entrance region of the charcoal beds plugged periodically; the entrance pipe (1/4-in..schedf40)'enters the bed normal to the section where the stainless steel wool is packed, thus the cross-sectional flow area (trap- ping area) at the entrance to the stainless steel wool is only 0.1 sq. in. | Although a particle trap was not installed between the second hold-up volume and the charcoal bed, it was considered and was definitely needed. Also the entrance section of future beds should be redesigned such that gas enters the empty chamber above the steel wool and thus offer a larger cross section of stainless steelrwool to gas flow as in the MK-II trap.. i 'Anothef argfiment.that'fiSSionlpfodficts, as well as hjdrocarbbhs, are - involved in the plugging mechanism at- the charcoal bed entrance, is the fact 'that the beds remained free of plugging until power operations were begun. - For a better understanding of the plugging mechanism, the MR-TI particle trap and the entrance seetiOn_tO'one of the charcoal?beds would need to be ekemieed. Also at least one bed could be examined along its length to de~ ~termine the fission‘produet'adsorption characteristics of the bed. The probable reason that the'ovetflow tank exit line remained clear for such a long time was the fact that pressure differentials (&3.5 psi) 214 were periodically released through the line each:time salt was returned to - the pump bowl. The pressure pulses probably cleared any incipient plug -which may have been in its formative stage. - The restriction at the fuel pump bowl does not seem to be related to fission products since the plug seems to form as readily if not more:so while subcritical and at low power as it does at full power. The gas exit line from the pump bowl should be modified to eliminate mist or liquid car- ‘ryover into the off-gas stream. A heated cyclone~type separator’? or a -large well baffled and cooled region which could be later heated to melt down any salt formation are two possibilities. o I1f hydrocarbons had not.beenrpresent°in_the pump bowl, fission gas be- ‘havior in the MSRE might have been somewhat different from that experienced. ;SometspeculationS'on-probable results are (1) less serious overall plugging because there would be no semisolid varnish-like buildup due to hydrocar- bons, (2) the York mesh would probably be less efficient at trapping the solid decay daughters and more of this material would then be trapped on the felt metal filter, (3) the fission product "cake' would probably be 1oose1y-packed on the filter and more easily disrupted by backflow since it would not contain a "binder" or paste material (hydrocarbons). 215 9. FUEL AND COOLANT PUMP LUBE OIL SYSTEMS | J. K. Franzreb 9.1 Description Two identical oil systems served tolubricate and-cool-the fuel and coolant pump bearings and to cool the shield plugs located between the bearings and the bowls of the pumps. Each was a closed loop designed to meet containment requirements, -The oil used was Gulf-Spin 35. Each system consisted of two 5-hp, 60-gpm at 160-ft head, 3500-rpm Allis Chalmers Electri-Cand pumps, (one normally in operation with the other in standby), an in-line Cuno. EFS oil filter, and an oil tank of 22-gal op- erating capacity- hav1ng "brazed on" water coils cap&ble of removing 41, QC0 Btu/hr. | | | . The oil flow to the bearings of each of the salt pumps was sabout 4 gpm. - with about 8 gpm to each shield plug. The remaining 48igpm was recycled through the oil tank to aid in cooling. A scavenging jefi_was installed in the o0il lines near the salt pump so that the shield plug oil flow aided the return of the bearing oil to the storage tank, Lubricating oil seeping past the lower shaft seal of the salt pump was piped to the oil eatch tank for measuring the rate of leakage. The pumps, Storage'tank,'filter and much of the instrumentation and valving for each system were mounted in an angle iron frame to facilitate moving to the site, This was'COmmohly referred to as a lube-oil package. The two packages were interconnected 80 that either could.be used in an ~emergency to supply both salt,pumps. - 9.2. TInstallation and Barly Problems - Before installation at the MSRE, both lube-o0il packages were operated ."in'a'test stand., Heafi‘loadiafid'preSSure drop data were obtained and the - performance of the sysfiems;was;eheeked.' A problem of gas entfainment_wae - found which affected the'prifiingzof the standby pumps. Approximately 30 seconds were requiredxto,fiiime'a gpare pump if it had'notrbeen.operated for several hours. Priming time was reduced to 5 sec by installing gas 216 vents from the pump volute casing and the pump discharge line to the gas space in the oil tank. This gxperiencé led to é modus operendi of starting the spare pump of each packagé once per shift and running it for about 15 min. . S : The packages were installed at the MSRE in 1964, Initial operation -was hampered by failure of the stator insulation in one of the pump motors and by low resistance to ground in stators of others. Since moisture was suspected, a potting compound (Dow Corning Silastic RTV-T31) was used to seal the motor hofising Joints and & moisture resistant coating of paint (Sherwin Williems epoxy white B69W6) was applied to the exterior surfaces of all four operating motors, plus two spares. , , ‘The loss of prime of the standby pumps continued to be a problem until: ’_the scavenging jets used to return oil to the reservoirs from the salt pumps - were replaced with jets of lower capacity to reduce the entrainment of gas - in the return oil streams. These replacements were done in September 1965, After modifications, it was possible to reduce the frequency of priming of - the standby pumps to once a week. Becsuse of the lowered capacity of the new return Jets, it was found necessary to limit the flow of oil to the fuel-pump motor to 4 gpm, as a flow of 5 gpm would result in a buildup of o0il in the salt pump motor cavity. ' 9.3 Addition of Syphon Tanks to the Qil Catch Tanks The oil catch tanks were febricated of & L6-1/2-in.-long section of " 2=-in. pipe topped by a 20-in;—longrsectibn of 8-in. pipe; The lower sec- tion allowed accurate measurement of the oil accfimnlationjrate whereas the upper portion provided sufficient volume to handle possibie gross:le&kage. The catch tanks were drained periodically to keep the.level in the lower. section. Radiation would not permit doing this during power operation and since the volume of the lower section was around 2500 cc and the allowable. seal lesk rate was 100 cc/day, it was possible that the reactor would have _to be shut down to drain the oil catch tanks during extended runs. There- fore, in October-1965, equipment was installed to automatically siphon .the _0il from the catch tanks when they became full. 217 After instellation, tests were run on these and they performed well. - However, they failed to function properly at the low oil leaskage rates that actually occurred during operation. The oil flowed over the high point of the syphon tubes, as over a weir, without bridging the tube to form a sy- - phon. We therefore reverted to manfialrdraining during shutdowns. Fortu- nately the leak rates did not get hlgh enough to interfere with operation of the reactor, 9.4 0il Leakage Continuous records of‘oilntankrandwoil catch tank levels were main- ° tained during all of the MSRE'operoting life to determine what losses were takifig place. Samples removed from the sysfems were'carefully measured, as were additions. ‘The oil storage tanks were large and therefore small . changes in level 1ndicatlon caused large errors in 1nventory. Successive ‘ log-readings could vary by 500 to 600 cc. Fairly accurate indication of ' the leskage through the rotating oil seals of the salt pumps to the oil . seals of the salt.pumps to the oil catch tanks were possible over one month or longer periods. The lesk rates during operation varied from a few cc . -per day for either salt pump to around 25 cc/day. The seals seemed to get "worse or improve for no apparent reason. For the 24-month period through August 1969, inventorles showed unac- counted for losses from the fuel pump oil system of 5.4 F 1.5 —3.0 the coolant o:l.l system of. D 6 4% (5) liters or a.vera.ge da.lly losses of 7.5 cc liters; from - (5.25 gms), and T. 8 cc (5.4 gms) respectivelyu Anelyses indicated that . there were approximately 1 to 2 gms of oil products per day in the fuel _;pump off-gas stream. ‘Some oil- nmywhave been held up in the lines, 6--f“t3 ':Fholdup volume, and. partlcle traps that were upstream of this sampllng point. In conclu51on, the best estimates showed an oil loss of 5+ gms/day, ,'of which only l-2 gms could be found as hydrocarbons in samples of off- gas from the primary system,_;e&v1ng approxlmately:3 gms/day unaccounted '?efor. 218 9.5 Change-Out of Oil Pumps . ‘After the initiel trouble with the electrical insulation due to mois- ture, the pumps gave very satisfactory service. Two pumpS'fiere removed from service, one because of excessive vibration and the other because of ‘an electrical short in the motor winding. This was done during the six- month period ending August 31, 1967. 7 The pump with excessive vibration had one of the two balancing disks loose on the shaft. This loose disk was reattached, and the shaft assembly was dynamically balanced. The pump was reassembled, tested under operating conditions and made ready for service. The pump with the shorted winding was rewound and put back into service. 9.6 Test Check of One 0il System B - Supplying Both the Fuel and Coolant Salt Pumps ~ On March 20, 1966, the two oil systems were valved so that the fuel oil package (FOP-2 running) supplied oil to both.saltrpumps; This was done to check calculations; and to be sure that this could indeed be done in some future emergency. The coolant pump was circulating salt at 1200°F; the fuel pump was idle and at a‘temperaturé of 125°F, Adequate flows to both salt pumps were maintained (3.35 gpm to the bearings and 6.5 and 7.3 gpm to the coolant and fuel pump shield plugs). Under these conditions the oil supply temperature equilibrated at 127°F. When the fuel pump oil flow was stppped,_thé oil supply temperature increased ~ T°F, indicafing that the fuel pump was removing part of the heat load of the oil system. Although this test didAnét,prove conclusively that one o0il cooling system was adequate for an emergency wherein one package would be used to supply both_operating salt pumps, subsequent heat exchanger tests and cal- culations indicated that one oil cooling system would be adequate,’ 9.7 0il Temperature Problems The total oil flow from one lube oil pump was normally 60 gpm. Of thig,.B gpm_was directed to the shield plug of the salt pump in the par- ticular system, 4 gpm to the salt pump bearings, and the balance was by- passed back into the oil tanks and caused to flow down khe outer tank walls. 219 This amount plus the warm oil from the shield and bearings was cooled by means of water flow through coils_brazed_to the outside of_theloarbon steel tanks. | . _ o o o | \ As the system was eubjected to extended service, scale built up on the inside surfaces of these coils, causing a drop in_the cooling capacity. This in turn caused the temperature of the supply oil-to climb from its normal 134°F to 140°F. This higher temperature was reached during March of 1966. R ) Both of the lube 011 tanks coils were flushed with a 15-wgt % solution of acetic ‘acid. Considerable material was removed and the 0il tempera- tures, upon: restarting, were reduced T°F in the case of the coolant pump oil system, and 3°F in the case of the fuel-pump o0il system. During subsequent operation the temperature of the fuel pump oil grad- ually rose, and in August. 1966 this acetic acid treatment was again given both coolers. This problem of gradual foullng of the water side of the cooling coils recurred throughout subsequent operations, and by December 1967, the tem- perature of the oil supplied to the pumps was up to 150°F. The water sup- ply was changed from tower water to cooler untreated ' 'process water" which in the case of the MSRE was taken directly from the potable water system via a backflow preventer. The system'was left on process water for two months and the heat'transfer_improved enough, probably by phyéical flushing out'offithe18cale,'so that tower water could again be used and afford satis- factory cooling of the 011. This suitching had to’he.done at least twice more before the reactor was finally shut down in December 1969 A semi~- '-permanent liose and piping system.was,prov1ded S0 thls:could be done expe- ‘ditiously. - _9,8 Increase in Radiation.Level at the Lube 0il Packages At thedbeginningrof]powerioperation iniAuguSte1969, a radiation moni- ‘tor at the reservoir of the fuel pump oil system indicated radiation from the upper part of the tank waerincreasing and'decreaSing'with reactor power. :U31ng the on-site gamma spectrometer, it was determlned that the activ1ty was qlArgon. This was. apparently produced by activation of the blanket gas 220 'in the upper part of the fuel pump motor cavity. Prior to this time only ‘helium had been used during power operation, and no gas'actiyation had oc- curred. Argon was being used at this time to investigate bubbles in the fuel loop. Since radietion levels did not ‘exceed 2 rR/hr and decreased when theé cover gas was changed back to helium, this was not a serious - problem. 9.9 ”Analysis of 0il Samples were taken of each new supply drum and from.the two lube o0il systems after circulating for a short time. Analyses of these were used as .controls, to compare with periodic samples taken from the systems during operation. The main concern was that the oil, especially in the fuel pump system, might undergo some changes due to long-time circulation through the .high flux field in the pump or that there might be some thermal decompo- sition. , ‘ o Table 9.1 is a compilation of some typical oil analyses performed during the life of the MSRE. The only significent or even minor changes that were found were that the OH radical 1ncreased somewhat and some C=0 was formed, probebly indicating some small degree of oxidation.__Tritium buildup was insignificant. | | The- oil which leaked through the seals to the oil catch tanks al- though dark_in-color, showed no significant chemical,or physical changes. An addition fractional distillation test analysis was run on & sample of new Gulfspin-35 oil on September 2, 196h This wes distilled under_re- duced pressure (3—4 mm Hg) in an 18-in. reflux colum. As indicated below, | most of the distillation occurred between 148 and 178°C | | 9.10 Replacement of 0il The first charge of Gulfspin-35 oil was added in September l96h as the 7011 tanks were calibrated Each tank took about 35 8allons.= Periodlc ad— -ditions were made to compenseate for losses. _ | As 1ndicated previously, no signiflcant physical oL chemical changes occurred which would indicate that the oil should.be replaced., Various oil t.companies were contacted and they did not have any suggestions as to other 221 Table S o . Viscosity -~ Viscosity Viscosity MSRE MSRE Date | Centistokes SsU SSU Carbon Sample No. Run No. Sampled Description = At 25°C At 100°F - At 210°F z L0-1 thru LO-5 | \ L 15.44 ‘ - 85.52 and o _ 5 ea 55-4al drums to . ’ s : to L0-9 thru LO-13: 8/25/65 new Gulfspin-35 - 15.99 - - 86.43 10-6 o 5 2/18/66 FOP-CONTROL NOT ANALYZED Lo-7 s 2/18/66 COP~CONTROL NOT ANALYZED L0-47 | | - 8/8/66 New 0il for FOP o 9.94 2.64 86.88 10-49 - 8/13/66 New Oil for COP | _ ' L0-69 11 2/13/67 From COP N 9.94 2.6 86.21 L0-70 11 2/13/67 From FOP , , 10.26 2.67 86.6 L0-140 5/16/68 ** . NOT ANALYZED - L0-141 5/16/68 %kx NOT ANALYZED ' ' _ Solids - ppm o LO-154 : 5/19/69 From FOP 180 9.94 - 2,58 78.0 LO-155 5/19/69 From COP 180 10.52 2.67 76.7 References: MSRE Sample Log for Water, Helium, Miscellaneous; and MSRE Data File — Section 6E-1 {Lube O: ek : : 01l in service from 5/20/67 to 5/16/68 — Collected in the Waste Oil Receiver from the fuel pump oil cat Ak 0i1 in service from 5/20/67 to 5/16/68 — Collected in the Waste 01l Receiver from the coolant pump oil ¢ -- TYPICAL ANALYSIS* .1 MSRE GULFSPIN 35 LUBE OIL | i o : H,0 Sulphur Bromine Flash Sediment SPECTROGRAPHIC ANALYSES -- Micrograms/ML ‘ N . % Number Point 2 viv Al Be Ca Ce Cr Cu Fe La Mg Mo " Ni P Pb Sn Sr T4 V Zn Zr Comments and Miscellaneous 9 1;26 3%3 F . <0,04 (all) ANALYZED AS ESSENTIALLY THE SAME AS LO-6 and LO-7 1.95 325°F <} <«0,02 124 <10 0,3 <0.4 0.2 <2 0.1 <0.4 42 74 <20 <20 0.5 <0.6 <2 96 <0,6 <1 <0.02 352 <10 0.5 <0.4 1.9 <2 0.2 <0.4 fiz 250 <20 <20 2.5 <0.6 <2 136 <0.6 ’ Mn Ha0-ppm , , 0.012 1.26 111°F 160 ? however, the MSRE and sampler were op- "erasted for 20 runs over a period of 5 years. 18.1.2 Eiperience with the Fuel Sampler-Enricher During Pre-Critical and Low-Power Operation i The sampler-enricher was installed and put in operation in May 1965. Sampler testing, shakedown, and operator training sessions were conducted concurrently with.the sampling and'énriching operations of Run 2. During‘ ‘this period, 53 fuel samples were fakeh, 87 enrichments made, and 20 oper- ators trained in the use of the sampler. Some of the problems encounfiered during this period were: 1. Buffer gas leaks at the operational (OV), maintenance (MV), and removal (RV) wvalves. ' 2. Failure of a solenoid valve on the removal valve actuator. 3. Failure of the removal valve to close completely requiring manual closure. | k., Failure of the access port (AP) to close completely. 5. Mbmenfiary failure of cable drive motor on capsule retrieval. 6. Rupture of the flexible containment membrane (boot) at the menipulator. ' - T. Deformation of the manipulator arm and fingers. 8. Accidental drop of an empty capsule onto the operétional valve. These and additional problems encountered with the subsequent 86 sampling cycles made during the approach to power runs are discussed in more detail in the following paragraphs. , _ Buffer Gas Leaks at the Operationél and Maintenance Valves —Fthring the sampling operations of Run 2, leaks developed through the upper metal- to-metal seats of both the operationél and maintenahce valves. The oper- ational valve was subsequently removed; examination showed that a thin black ring, which was easily removed, had formed at the upper sealing sur- face of the valve gate and a small quantity of salt spheres (<1 gm) had 297 collected between the seats of the gate. Apparently salt particles were dislodged from the capsules during the operation where the capsules (after ~ having been dipped in the pump bowl) were disengaged from the latch above the valve. When the stem and~gste were lubricated, the valve sealed almost completely; however on removal of the lubricant, the buffer gas leak rate increased to 2 ce/min through the upper gate seal and remained at zero through the lower seal. A few sampling cycles after the valve was reinstalled, the leak rate again increased to 20 cc/min through the upper seal. Apparently more salt particles had lodged between the buffer-gas sesling surfaces. Repeated ef- forts to blow the particles from the sesaling surfaces failed. Since there were three other sealing surfsces between the pump bowl and the sample ac- cess area and since the leak did not increase with continued sampling, the valve was not replaced. | | | After approx1mately 86 additional sampling cycles, an opportunity to clean the operational valve arose when the reactor was drained to repair the cable drive motor in April 1966 (to be discussed later). Cleaning the seating surfaces decreased the leak rate for a while until an empty cap- sule was accidentally dropped on the gate. Subsequent valve operation again resulted in a high buffer-gas leak rate. The buffer gas seal at this surface continued to deteriorate slowly throughout the remainder of MSRE operations; the only consequence was a slow pressure buildup in the con- tainment volume sbove the valve which required periodic venting. From February 1967 until final shutdown, the buffer supply pressure to this seal was used only when the sampler was in operatlon thereby dbv1at1ng the perl— odic venting of area 1C. o ' o During the latter part of 1966 there had been a gradual increase in leakage of buffer gas through the upper seat of the maintenance valve also. - Although the procedure specifled-that the maintenance valve be opened be- fore.the operational-valve'is'ofiéned, and that the operational valve be “closed before the maintenance valve is closed, on at least one occasion the ‘reverse sequence was followed during closure and thus foreign matter could have been dislodged from the upper velve onto the lower valve, While the lesk rate remained relatively small (v25 ce/min), it was sufficient to 298 cause difficulty with proper operation of the interlocks (an electrical signal which indicated that the valve was indeed closed was based on the ability of seating surfaces to maintain a specified buffer pressure). Since '~ cleanup would have been difficult and replacement of the valve was not wér— 'ranted by the lesk rate alone, a mechanical.method of assuring that the valve was closed was sdbstituted.for the pneumatic methoda No additionsl problems with these valves were reported. Area 1C Access Port — The access port door was operated by six clamps vhich relied on pneumatic operators and a spring for opening and a sched- uled time-delay of these pneumatic operators for closure;lhowéver,'once closed, the mechanical linkage was such that the clamps no longer require ~gas pressure to keep the dobr tightly closed. The ability of twolconcen— tric neoprene gaskets (between the access port door and housing) to contain helium at 55 psia satisfied an interlock and indicated that a gas-tight seal did indeed exist between areas 1C and 3A. On several occasions during the pre-critical run, the access port failed to close properly. Also, one or two of the six access port operators ffiilgd to open and the manipulator. wvas used to release the sticking operators. During the shutdown following the low-power experiment, the clamps were readjusted and,an extension was added to the pin of each clamp; this extension made it easier to open any clamp manually with the manipulator in the evént of a pneumatic malfunction. Flexible Containment Membranes — There are two 0.020-in. thick flexi- ble urethane membranes (boots) which partially enclosed end provided mo- bility to the manipulator arm used in transferring capéules to and from the latch in area 1C and the bottom portion of the transport carrier in aresa 3A. A vacuum was maintained between these boots to .monitor the integrity of the boots and- assure that double containment did indeed exist between the sampler operator and the sample. A ioss of vacuum and a pressure re- duction in area 3A would indicate a leek in the outer boot whereas a loss of vacuum with no pressure change in area 3A would implicate the inner boot (atmosphere side) or possibly both boots. There were three boot failures durlng the first 1Lk0 sampllng cycles of Run 2. One was caused by an inadvertent evacuation of area 3A while the manipulator cover was off; the resultant pressure gradient ruptured the boot. On another occasion, the boot was snagged on the bottom piece 299 of the transport tube in area 3A when the manipulator was used to release -a stuck access port clamp. The third failure reSulted from pinching the - boot between the manipulator arm and the housing. ' At the end of the next run (Run 3), a pressure switch (with an alarm and interlock) was installed'to detect and prevent a negative pressure gradient greater than 1/2 psi between area 3A and the manipulator arm. Steel rings were installed in the convolutions of the inner boot to hold it free of the manipulator arm and p0551bly prevent the pinching-type failure. _ - Also to ensure thaf at-least two barriers existed.between.primary con- tainment and sampling persoanel,_a pressure switch was installed on the ‘manipulator cover which required & negative pressure of k-in. Hg in the cover before either the operational or maintenance valve could be opened. | In Aprll 1966, the manipulator cover pressure was inadvertently evacu- ated to v25-in. Hg w1thout 1ower1ng area 3A pressure simultaneously; the 12-psi differential caused a small puncture of the inner boot. This set of boots had been used for 48 sampling cycles.eince,power operation was - begun and the maximum power_reached at that time was 2.5 MW. The radiation level of the manipulator assembly and boots was 10 R/hr at 3 inches from the fingers, but was reduced by a factor of 10 by scrubbing with soap and water. About 2 hours were'required for replacement of both boots. Deformation of the Manipulator Arm and Fingers fi—-Durlng Run 2, the manipulator arm and. fingers were bent which causedrdlfflculty in gripping ~ the latch cable and moving the manipulator arm. At the end of Run 3, the arm was replaced and & l/h X l/h inch projectlon vas welded to the bottom "of each finger to aid in grasplng the capsule ceble from the floor of area 3A§ The manlpulator arm was replaced and the clearances between the arm and castle 301nt were increased to reduce the force requlred to operate ~ the manlpulator. The manipulator and fingers operated satlsfactorlky for 421 addltlonal - sampling cycles until August 1968 at which time the manipulator assembly - was replaced because the tips of the fingers no longer closed tightly. The old assembly was decontamlnated and repaired. The replacement fingers op- erated satisfactorily untll September 1969 when the double metal bellows 300 deve1oped a leak, The sampler on the fuel processing System'was canniba- lized to make the repairs; this unit operated satisfactorily for the sub- sequent three months of reactor opersation before shutdown. ~ Capsule Recovery from the Qperational Valve — On one occasion during the precritical‘run; an empty sampling capsule was accidentally knocked into area 1C before the latch pin was completely engaged in the latch, and the capsule dropped onto the gate of the operational valve. The capsule was retrieved by rémoving the manipulator assembly from 3A, opening the access port, and snaring the capsule cable with a wire hook. Since this type of accident could reoccur, the brass latch pins used to attach the capsule to the latch were replaced with nickel-plated mild steel pins so that capsule recovery could then be effected with a magnet. After approximately a month of full-power operation in July 1966, an empty capsule again was dropped onto the gate of the operational valve when thé manipulator slipped during the latching operation. A magnet'was low- ered into fihe transfer line and the capsule assembly was recovered as the magnet was withdrawn. The radiation level of the magnet after the recovery operation was 10 R/hr at 2 ft. All sampling capsules of the bucket-type assembled since the latter part of 1966 contained a nickel-plated mild steel top so that capsule re- covery could be made with a magnet. Removal Area, Valve, and Seals — During the pre—critical run, the removal seal and valve required realignment and increased tolerances be- fore the transport container and removal tool,assembly,would.slide through these units freely without binding. In addition; the,rémoval valve failed to seal properly even though the ball and seals were replaced, Therefore, at the end of the next run, the.valve assembly was replaced with a modified version to improve the sealing characteristics and improve future access to the valve. This unit served satisfactorily except that a buffer gas leak developed in the top teflon seal of the valve dufing the latter part of 1967° As a consequence of the leak, the equilibrium buffer seal pres- sure feached only 30—35 psia instead of the normal 50-55 psiaj; the associ- - ated pressure switch which pefmits the opening of the operational and.main— ‘tenance valves and the access port was lowered from-SO to 30 and finally in December 1967 to 25 psia where it remained for the remainder of reactor operations. 301 Latch, Capsule, and Csble Malfunctions — Early in Run 2, the cable - drive motor stalled during the withdrawal of an empty capsule. The cap- sule was reinserted sbout .12 ih. and then withdrawn successfully. The reason fbr_thermalfunctioo;fias unknown., Approximately 100 sampling cycles later in December 1965, the @otor stalled again while withdrawing a 10-gm sample. After repeafied attempts, the capsule was withdrawn completely and " the cable was examined., The capsule~was5empty_and»the'cable_was found to be backed up into the drive'finit box and caught in the motor gears. It was assumed. that the latch had hung on the gate of the operatlonal or main- tenance valve causing the. cable to coil up inside area 1C, The cable was straightened, the limit switches on the operatlonal ‘'and maintenance valves ‘were'reset_to_open;the velves wider and the diameter of.the latch was re- duced. . " In April 1966, after approximately a week at a power level of 5 MW, - the capsule drive-unit motor stalled as the latch was being retrieved through the pipe bend near the pump bowl. After several inserts end with- drawals, the latch was retrieved. The latch and part of the cable were T_then-pulled through the access port, the removal valve, and into the sample transport cask which provided shielding while the latch was replaced with | one designed torprovide-anfadditional‘l/8—inch'diametrical clearance, Approximately 3 months laterithe capsule stuck again temporarily for some unexplained reason on withdrawal and no sample was obtained. Subse- 'quent hangups - with ‘more serious consequences are discussed in Section 18.1.3. Repair of an.Open Electrlcal Circuit — While testing the new latch, the cable position, 1nd1cator stopped ‘when the latch was approxlmately two feet from the pump bowl* the;upper'limit svitch also actuated at this time, ',Electrical continuity'checks showed open. circuits to the‘insertand with- 'n;,drawvwindiags'of the cable:drive'motor and also to the upper limit switch. Allgthese-leads-pehetrate:fhe7con£ainment wall through a'cofimon 8-pin re- '”ceptacle. ‘Since repalr could not be made while fuel was in the reactor _Jl;without violating contalnment the ‘reactor was drained. The cable and latch were then pulled into Area 3A using the manlpulator to pull and the : transport container and access port 0perators to hold the cable’ so that the cable could be regrasped w1th the manlpulators for another pull The 302 ~ operational and maintenance valves were then closed. The assembly includ- ing thé drive motor, cable, and latch were removed, partially decontami- nated, and repaired with the use of shadow shielding techniques. (Three connector pins in one of the 8-pin receptacles were burned off.) All six similar receptacles at this location were filled with epoxy resin to ‘pro- ‘vide additional insuletion and strength., The cable which was bent in sev- eral places during retrieval was decontaminated with soap and water to_re— duce the radiation level to 5 R/hr at 3 'in. before it was straightened. After reassembly and checkout, normal sampling was resumed. ' ' Contamination of the Removal Seal — An area at the top of the Sampier - was provided so that the lower part of thertransport carrier and capsule could be evacuated and pressuriied several times to remove oxygen and mois- ture from the capsfile and carrier before insertion into the dry box (Area 3A) of the sampler. The bottom seal for the removal areaiwaS‘provided by the removal valve and a seal at the top was made when the transport tube was inserted through a set of O-rings at the top of the.rembval-volume; During sampling the shipping cask was aligned and placed over the removal area, A small amount of vacuum grease was smeared onto the lower part of the transport tube before it was lowered through the shipping cask and into the O-rings of the removal seai° When the transport tube was further lo- wered into Area 3A (a contaminated area) to insert an empty capsule and again to pick up the sample, the ocuter surfaces of the transport tube and removal tool became contaminated with solid fission products from the cap- sule, floor of Area 3A, and/or manipulator. Some of these particles were wiped off by the O-rings of the removal seal as the removal tool and trans- port tube were withdrawn from Area 3A; that remaining on the removal tool was wiped off with a damp cloth. The shipping cask was gradually contami- nated as both the removal tool and transport tube were drawn ifito it, By this process the removal seal becomes progressively more contaminated with each semple. During pre-critical operations this transfer mechanism for particulate matter was not obvious; however, after power operations had begun and especially after the latch maintenance work, where the latch was manually retrieved from +the vicinity of the pump and out thrbugh the re- - moval aree for repair, the removel area and the fop of the sampler became contaminated, > 303 * The adoption of the'folldwing procedure'proved effective in preventing ~the spread of contaminatlon durlng sample removal and transport throughout ~ the remainder of reactor operatlons" " (1) The top of the sampler was establlshed as a Contamination Zone. (2) The removal tool was wiped with a damp cloth durlng withdrawal from the removal area. 7 i - (3) The shipping cask was wiped with a damp cloth prior to enclosure in a plastlc bag for shlpment to the analytlcal laboratory. (%) The top of the sampler was wiped with & demp cloth after each sample. | _ - _ (5) An exhaust hood was erected near the removal area and.partially enclosed the shipping cask'to'maintain & slightly reduced pressure in this ~area so that any airborne particles would be drawn into the fllter and ex- haust system of the bullding. (6) The removal seal" area was cleaned periodically with damp wipes. - Control Circuitry Changes ——-Durlng this perlod five changes were made ' to the sampler control clrcults ' (l) A pressure switch was sdded to prevent evacuation of Area 3A to ‘more than 10 in. of water greater than the manipulator cover aresa. (2) An interlock was*added'to'require that both the operational and _'maintenance valves be closed before the access port can be opened. (3) A permissive sw1tch and light were installed to indicate that the access port can be opened only when Area 1C pressure is equal to or "less than Area 3A pressure.a_,_~,f-,',, , | | | (4) A fuse was added to the capsule drive motor circuit to protect 'the motor and the electr1ca1 receptacles from excess1ve currents. ' (5) Voltage suppressors were placed across the two motor W1nd1ngs rto limit any high voltage peaks durlng startlng and stopplng.,- ' o Mlscellaneous Sampler—Enrlcher Prdblems_-—-Durlng the 1ow—power experl- ',dment a five-inch—long capsule W1th an opening nesar the top and capable of ~a1hold1ng SO-g of salt was lowered 1nto ‘the pump bowl but falled to trap a _“sample because the assembly was not long enough to permlt total 1mmer51on. -Subsequent longer assemblies (9) obtained salt except one which was believed "to have hung on the latch stop ‘at the pump bowl and did not enter the pump. The capsules were subsequently modified to allow them to hang straight. 304 *Most of the samples taken after the reactor exceeded 1 MW (nuclear power) were highly radioacfive and exceeded 1000 R/hr at 3 in. Fission products adhering to the outside of the capsfiles contaminated the,Bottom cups -of the transport containers. For a limited time disposable plastic liners inserted in the cups were effective in reducing the radiation level in the bottom portion of the transport tubes. However, as more fission 'produets built in the fuel and the interior of the sampler became more contaminated, it was more economical to fabricate disposeble,bottom,cups of mild steel rather than decontaminate and reuse the stainless steel'eups° The new cups were shorter than the previous ones and the plastic liner was used to confine the capsule, wire, and latch pin so that the top part of the trahsport carrier could be more easily slipped-over this combination to engage the threads and double O-ring seals at the bottom of the mild -steel cups. | , | On one occasion while the top of the transport containeriwas being mated with the bottom which contained a 50-g capsule, the wire on the cap-. sule caught in the threads of the matingfpieces and galled before the two pieces were sealed completely. Thereafter the liners were lengthened to accommodate the the 50-g capsules., In July 1966, the fuel pump was accidentally overfilled with flush salt and salt was pushed up into the sampler tube approximately two feet above the fuel pump where it froze and thus prevented the capsule and latch 'from being lowered into the pump bowl. The salt was melted and allowed to drain back into the pump by energizing a set of remotely-placed heaters around the sampler tube ‘and the normal heaters surroundlng the pump bowl. Normal sampling was subsequently resumed , Contamination of the area surroundlng the sampler was mlnlleEd during malntenance by designating the area surrounding the sampler a Contamina- tion Control Zone and at times. by erecting a plastlc tent around the sam- pler. Repair work was done inside the zone with approprlate protectlve ciothlng, gloves, and shoe scuffs. Contamination was found outside the - zone twice and on_both occasions it was attributed to contaminated shoes which indicates that this type of contamination is not readily airborne. 305 18.1.3 Subsequent Operating Experience with.the,Sampler—Enricher- | | The major sampler problems ‘encountered durlng the remainder of opera- tions were. caused by the latch and/or capsule lodging at the entrance to the sampler tube or at one of the isolation valves 1mmed1ately below Area 1C with subsequent unreellng and tangling of. the cable 1n Area lC and in ‘bhe gears of the drive wnit. , Ruptures or pln-hole leaks in the boots enc1051ng ‘the manipulator arm were a chronic source of troublefrequiring perlodic replacement., Life of the boots varied videly but avenaged L) sample cycles before failure. The cause of most of theserleaks:was undetermined because the boots could not be examined easily due to therhigh radiation level of the manipulator as- gembly, | . - | | Gradual deterioration of the buffer gas seals at the operational, malntenance and . removal valves and at the access port continued and on occasions the time required to take a semple was lengthened because & longer time interval was required for the buffer pressure to build up suf- 'ficiently to actuate the proper.relay for the next step inithe procedure., Loss of Capsules in‘fhe“Pump Bowl = In. ‘August 1967 during what ap- peared to be 8 normal sampling operation the capsule and/or latch had ap- parently lodged ‘at the maintenance valve while being lowered causing the icable to coil up in Area 1C and into the motor drive area where it had be- 'come,snarled and kinked. The first indication of a melfunction was during the capsule retrieval cycle when the motor stalled_with-B‘ft'3 in. of cable off the reel.53 Repeated inserts and withdrawals recovered only 5 more in. of cable before the motor stalled with only &n inch of travel in either di- ";)rection (thls had happened once-hefbre in 1965 but ‘the cable and capsule ~ ‘were-retrieved completely at that time) | | . If the capsule and cable were lodged at the sampler tube entrance be- :“'low Area 1C, then. the operatlonal and maintenance valves could be closed ;and sampler repairs could proceed without a reactor drain, otherw1se a drain would be necessary. After con31deration of several poss1b111t1es, it was decided to disengage the motor drive from the operational valve and close the valve slowly with the hand wheel W1th the idea that if the cap- sule were below the valve, the cable would offer enough resistance to clo- 'surerto be detected in which case the valve would be reopened and the 306 reactor drained before repair. No resistance was detected until the valve was practically closed. Therefore,'bcth valves were closed and when Area 1C was opened, as expected, the cable was coiled up within;_hekever the latch, capsule, and sbout 6 in. of cable were;missing.“The reecfior was then drained and grappllng tools were fabricated of flexible tublng and cebles which could be inserted through the sampler into the pump bowl., Four of the five tools which were most successful at grasping a dummy latch are shown in Fig. 18.3. _v | ; - When'the.noose tool was'lowered, it&snared-the latch but the 1/32-in. steel cable broke because the latch was apparently glued in place with salt at the latch sfopu‘ The_dislodgihg tool was then fised but apparently to no ~avail because & second noose grasped the laich but could not budge it with a pull of 25 1b. The pump was then heated until the latch was at a tem- 'perature of TOO°F. It was then 1ifted several feet without difficulty be- fore it lodged in the tube, apparently at the expansion joint between the fuel fiump and the sampler. A corkscrew tool was next inserted but the latch again hung on retrieval this time at the first bend near the fuel pump. The latch retrieval tool which was de51gned to fit over the severed cable and stem and thus prevent the ceble from jamming against the wall ‘during retrieval was successful in pulling up'the‘csble;’letCh,'snd latch pin. The capsule apparentlylsnapped_its-cable and dropped to the bcttom of the sampler cage when the latch ceme to & sudden stop at the latch stop. The dotted outline of & capsule in Fig. 18.k4 shows how & capsule, once de- ‘tached,from‘its c&ble, could slip out of the sampler cage and then be trapped by the mist shield in the pump bowl. - Bince the capsule top is made of niCkel plated mild steei, a magnet in combination with & go-gage (Fig. 18.3) was lowered into the pumphbcwi to verify that the sampling tube was unobstructed end to pick"fip the cap- sflle; Capsule recovery was unsuccessful and'since no adverse chemicsal or mechanical effects were envisioned should the cepsule be allowed to remain- in the mist shield, further retrieralrattempts were Ebendoned While retrleval efforts were in progress, & replacement 1-C assembly vhich was on hand was equipped with e latch made of 430 stainless steel and with a sleeve to prevent future tangling of the cable in the gears of the drive unit. However, subsequeht examination of the 0ld drive unit in 307 ORNL-DWG 67-1788 Y/g-in CABLE -~ | i ~'/4=in. CABLE i/32-in. CABLE tYa-in. DIAM Vo ~in. FLEX CABLE SHEATH CABLE BEING NOOSE READY PUSHED OUT FOR FISHING NOOSE TOOL ’ DISLODGING TOOL _ ine ~in. CABLE / 'fls ~in. CABLE Uz 4 ’ tin. 1D | / COPPER ’ ' a |1/, in. SCHED 40 PIPE % [4 | 3,in. DIAM x 3%2in. [ . MAGNET—- ; / : M ) ; 5 TE ~1.375 in. DIAM - BRASS T " LATCH RETRIEVAL . ... . GO- GAGE AND TOOL CAPSULE RETRIEVAL = o TOOL o Fig. 18 3 _Tools Developed for Retrieval of Latch and Capsule from s Sampler—Enricher 308 ORNL-DWG 67-10766 LATCH ASSEMBLY CABLE SHEARED OFF LATCH STOP. FINAL CLOSURE WELD T A e T L L TN T T TLs LT oYy VA A A A A S A A A LS N R B, T TR T T T T R T g een p il | T Fodndd T NORMAL OPERATING SALT LEVEL CAGE FOR SAMPLE CAPSULE Lt T T TN s SsS ST TSNNSO, — A ACTEITTCYS SALT INTAKE SLOT Fig. 18.4 Location of Sampler Latch and Capsules in Fuel Pump Bowl APPROXIMATELY AT THIS LOCATION LATCH ASSEMBLY NORMAL POSITION TOP OF FUEL PUMP = P ,v.‘.': — BAFFLE POSSIBLE MOVEMENT OF CAPSULE SAMPLE CAPSULE SAMPLE CAPSULE CABLE e ; SECTION A-A SAMPLE CAPSULE LOST SAMPLE CAPSULE 3/ in. MAX —‘J: 1 0 ' 2 3 4 ,(;&‘ INCHES 309 a hot cell revealed that the cable waslnot tangled in the gears as had been assumed; the failure to withdraw or.insert was due to a sharp kink in the!cable-whicfi beceme lodgedrin the 17/32-in. diameter channel of the - latch positioner. | . When normal operations and sampling were,resumed, s commercially avail- sble device,was.testedrfihich was capable of sensing the latch (now made of magnetic material) inside the sampler tube. A design was prepared to mount this proximity switch about 10 in. below the maintenance valve to detect the latch as it is lowered during normal sampling; however, before it could be installed, the latch agaln lodged in the sampler tube without detection in March 1968, The usual 17 ft 5 in. of cable was reeled off and was tan- gled in Area 1C. As it was rewound, the motor stalled with 13 ft 5 in. off the reel. When repeated withdrawal attempts failed, the reactor was drained and the capsule access portrwas_opened._ The capsule was visible through the rlcwerecornerwof the opening and the latch could be seen in the back of the chamber w1th many loops and coils, some of which appeared to extend down 1nto,the sampler tube. Rather than risk cuttlng the cable the isolation valves were left open; however, when an attempt was made to 1lift the cap- - sule out.of the chamber, the manipulator brushed some of the coils causing them to spring out through the port and pull the capsule wire from the ma- nipulatorsfingers. The capsule dropped into the sampler tube and the latch tipped over, thus releasing the key; the capsule and key_thefi disappeared down the sampler tube in what appeered»to be a bottom-up position. After. several unsuccessfulrattempts were made to retrieve the capsule by lowering magnets of varlous s1zes into the pump bowl, the reactor was filled with flush salt. A 50—g capsule (5 in. long with an opening 4-1/8 in.efrom the bottom of thejeapsule)_wes lowered 1nto_the.pump.butmcame up empty; salt dropletsrclinging”te.the outside indicatedrthet'the capsule had f'beefi.enly half sfibmefged Later a 10-g cepsule did collect a sample. In the meantime a full-scale plastlc and metal mock—up of the sampler 'tube mist shield, and sampler cage was constructed and used to select the . best tools and magnets which could be lowered via the sampler in another attempt to retrieve the capsule,sh The simplest and most effective tool proved to be 1/2 and 3/4-in. diameter Alnico-5 magnets. When the flush salt was drained and retrieval efforts resumed, sounds. from a contact 310 microbhone on the fuel pump indicated that the 1/2-in. dia. magnet lifted an object & few inches and then dropped it. After many sttempts, the mag- net came up with an obJect'which 1ater proved to be the corroded top of the old capsule. Further efforté-wiéh_either magnet were unsuccessful and the second capsule was abandonéd. | ‘ ' During this shutdown, the proximity switch was installed sbout 4 in. below the maintenance valve and the-sampling frocedure was modified to stipulate that if the switch is not actuated when the capsule position indi- cator shows that sufficient cable has beqn-réeled off to reach the proxé imity switch, the drive will be stopped. ‘' Also a variac was added to the drive motor circuit SO'that'the'operatihg voltage‘éould be éhanged’to 80 ~ volts to lessen the possibility of seriously kinking the_éable; - | - During the next stertup, with flush salt in the reactor, salt was " trapped in a 10-g capsule but none trapped in a 50-g capsule which requires an immersion depth of 4-1/8 in. to trap & sample,-'However;'é 30-g capéfile requiring only 2-1/2 in., of immersion did trap a sample. These results agree with the measurements taken during capsule retrieval which indicated that the abandoned capsule was 1-3/4 in. above the bottom of the cage. - However, with fuel in the resctor, the actual level in the pump bowl was somewhat~highér than that indicated because of the high void fraction of 'the liquid in the pump bowl.r Consequently no problems wereé encountered taking 50-g fuel samples throughout the remainder of operations eiéept' during the latter part of 1969 when the fuel pump speed was lowered to 980 rpm. At this speed the fuel salt void fraction is approximately equiv- alent'to_that*fdr flush salt at normal pump speed. The failure to trap & 50-g sample under these conditions indiéated that the positiofi-of the aban- doned capsules had not changed after approximately one year of operation. In October 1969, & serious ceble tangling problem was averted by the use of the proximity switch. When the switch failed to actuate after 3 ft 10 in° 6f cable was peid out during an otherwise normsl capsule insert, the cable was retrieved until the position indicator indicated that all of the cable was back on the takeup reel and the capsule was fully withdrawn into isolation chamber 1C. The operational and maintenance valves were then closed and the access port was opened. The cable was indeed fully re- trieved ag indicated and the capsule was fufily visible but was lodged 311 diagonally between the-ledgeé.of the doorway to the chamber. Apparently the capsule had lodged at the sampler tube entrance or at one of the isola- tion valves and as the ceble was paid out,‘it somehow encircled the capsule because the capsule was lifted and relodged at a higher position (during cable retrieval) than it was at the start of the sampling'procédure. The capsule was then lowered to its normalAposition_and all subsequent cable operations proceeded without interruptions. Wiring Fault,——-In.November of 1967 as a sample was being withdrawn, the 0,3-amp fuse in the dri#e'motor failed. Subsequent tests showed that the insert mode was normel-at 0.2 amp but during withdrawal, the current ‘increased to 1.0 or 1.5 amp, indicating leakage.to ground. During further tests, three_circuits opened, disabling the insert and withdraw circuits and the upper limit switch. After the erection of & tent and other meas- ures to prevent the spread of-contamination; a 3-in, diam hole was sawed in the cover plate of Area 3A directly above the cable penetration into ‘the inner box (Area 1C). The failure was found where the wires were bent back against the side of the plug on the lower end of the cable between the inner (1C) and outer (3A) containment boxes. When a temporary con- - nectionrindicated that everything within Area 1C was operable, the sample was retrieved and the isolation valves were closed. The damaged section | was abandoned and a new cable was installed having a penetration through 8 b-in. pipe cap welded over the sawed hole in the top cover. mWhile the pi?e'cap was being welded in place, a heavy current evi- dently went to ground through an adjacent penetration, destroying the plug . end receptacle. Repairs were made'by cutting out and replac1ng this pene- tration. As & result of thisgexperlenCe, an_isolatlon transformer and fuse '~was'added.to each of thétthrggtdrivefunit-cables which*allowsrdne ground ' without interference with obéfétion;[ After the repairs, all eircuits were roperable except the upper limit switch which stops the drive motor when the _latch reaches the latch stoP._ Since the motor and circuit can easily with- stand blocked—rotor current “the upper limit switch was bypassed and there- 7 after the motor was turned off when the p031tlon indicator indicated that the 1atch was fully w1thdrawn,u Repalr of Cable Drive Gear — In December 1968 during an attempt to remove & capsule from the latch, it was found that the cable could be 312 pulled off the reel vhile the drive motor was stationary. The tentative diagnoéis was that one of the'pair,of drive gears was slipping on its shaft. In order to galn access to the drive unit, it was necessary to re- move the shield blocks over the reactor cell which, in turn, required that the fuel be dréined. After a temporary containment enclosure and contami- nation zone had been set up around the sampler, the conteinment box, 1C, containing the drive unit was disconnected, lifted, and a 3-in. hole sawed through the side adjacent to the gears. One gear was found to be loose be- cause its two setscrews had come loose. The gear was repositioned and tack- welded on its shaft and the other gear was fastened withrjam'screws on;tdp of its setscrews. A patch was then welded on the box and the unit rein- stalled. The work was done without excessive exposure of personnei despite the high radiation levels (10 R/hr at 12 in. from the box) by use of shield- ing and extended tools designed especially for the Job. ' Since other parts of the sampler were easily accessible during this shutdown, the manipulator hand was repaired, the illuminator port and view- ing window lens were replaced because of discoloration by radiation, and an imperfect seal in the removal valve was replaced. - Vacuum Pump Problems ——-The_sampier was equipped with two vacuum pumps. Vacuum pump No. 1 was used to evacuate contaminated areas such as contain- ment areas 1C and 3A and discharged into the auxiliary charcoal bed; vacuum punp No. 2 was used to evacuate non-contaminated areas such as the manipu- lator cover and the plenum between the manipulator boots and discharged in- to the containment air system filters and to the stack. During the July 1966 shutdown, the oil levels in the pumps were checked and a small amount of oil waéfladded, _ In January 1969 there was an activity release to the stack (<0.08 mc) which was attributed to a gas lesk around the shaft of vacuum pump No. 1. The pump was replaced because the oil seal at the shaft could not be re- .paired without extensive decontamination procedures. Four months later, repeated vacuum pump motor outsages (due to overload) were apparently caused -by low oil level. On one inspection, the pump weas practically empty of oil with no evidence of external oil leasks and was consequently refilled to the proper level. On the next inspection, the pump contained approximately . 313 2 qts more than the normalhinventory. Apparently the previously lost oil drained back into the pump. High pressure (>13 psia) in Areas 1C was nor- mally vented into the.auxiliary charcoal bed through a line bypassing the pump. Apparently high pressure was vented through the vacuum pump instead, thus carrying over a large portion of the oil into the holdup tank and off- gas line above the pufip discharge. After the 0il level was readjusted, 'pump operation returned to ndrmal>and the last oil check was normal. Heated Shipping Cagk —- At temperatures less than LOO°F, radiation from fission products can produce free fluorine in frozen fuel salt, To minimize fluorlne production and thus avoid the effects of fluorine on the results of oxide and trivalent uranium analyses, a shipping cask was fabri- cated which‘maintained-the.fuel sample at approximately S00°F from the time -itrwas removed from the sampier until it was unloaded at the analytical ’iaboratory. The cask used molten babbitt both as shielding and as a heat reservoir. Built-in electric heaters melted the babbitt before the sample waS'loaded and the heat of fusion maintained the samplerat a relatively constant temperature for about 9 hours during shipment and storage. For the elevated temperature, O-rings made of Viton A were substituted for the standard neoprene O-rings in the transport tube. In addition to the double O-ring seal provided by the transport tube, the transport tube was capped inside the cask during shipment. All samples were unavoidably cooled to room temperature in the sampler where they were sealed in the transport tube at atmospheric pressure. When the transport tube was inserted into the heated cask, the pressure in the - tube 1ncreased to approx1mately 10 psig because of the temperature inerease. :‘On one occa51on when the cap was removed from the cask airborne activity was released from the cask caV1ty into the ventilation system of the un- 1oad1ng statlon at the analytlcal laboratory. Subsequently the sealing surfaces of the top part of the transport tube were found to have been eroded by repeated decontamination procedures to such an. extent that the double O-ring seals were ineffective. After examlnatlon of the sealing surfaces of the remaining five transport tube tops and reevaluation of the decontamination costs, it was;decided to discard the reusable_stainless steel transport tubes and use "throw-away" transport tubes made of mild 314 steel for all subsequent samples. A valve was later added to the'cap of the heated cask so that any future pressure buildup in the cask cavity could be released to a charcoal filter before cap removal. - Other Capsules and Samples — The size of the capsules,fhat could be accommodated by the samplér'wds 1imited to 0.75 inches in diameter and 6-1/4 inches in length. 'Approximately 1/4 inch greater diameter could be accommodated by the sampler but not by the transport'tubee Four of the various types of capsules which were lowered into the pump-bowl are shown in Figures 18.5a through 18.5d. The capsule in Fig., 18.5a was used to trap "either salt or gas. The capsule was first evacuated and sealed with a mea- sured emount of salt similer to the fuel. On entry into the fuel pump, the - sealing salt mgited and salt or gas (depending on the capsulé elevation in the pump bowl just prior to salt ligquefaction) was drawn into the capsule. | When the beryllium addition capsules were withdrawn, they were en- crusted with solids as shown in Figure 18.5b., As a result, the capsules were difficult to loed into the bottom part of the fransport tubes even though the plastic liner was removed. Also the enriching capsules were sometimes difficult to load because of their length (6-1/L4 in.). The ca- bles of these lohg capsules were occasionally caught in the threads of the mating parts of the transport container and were difficult to unload at the -hot,ceil.~r ' | L The.iadiation levels of the beryllium capsules and the empty enriching capsules fiere more than four times higher than the steandard 10-g sample when delifiered to the analytical laboratory. N . _ ' Miscellaneous Sempler Problems — During the latter part of 1966, 8 cotter key had come out of the top of the lower hinge pin on the access fiort door. The‘hinge pin had then worked out of the top of the:hi'ngeo Using only'the manipulator, the pin'was repositioned in the'hinge, 8 new cotter key was inserted in the pin and the key was spread to lock it in place. During this repair, the lowei key was dislodged and was also re- paired.- ' | - ) Operation of the pneumstic clamps on the access port door occasion- ally resulted in gaseous fission products being vented through the operator discharge lines. Therefore a small charcoal filter was added to the op; erator discharge lines to prevent this activity from'being discharged to 315 ORNL-DWG 67-4784 |_.—STAINLESS STEEL | CABLE VOLUME |_u—Ya~In. 0D NICKEL 2060 ——__ o | - NICKEL CAPILLARY. LiyBe F'4\ Fig. 18.5a Freeze Valve éapsule PHOTO 9L278 Fig. 8.5b Nickel Cage Beryllium Rod Assembly after Exposure to MSRE Fuel Salt for 10.5 hr b e i e bt ¢ MIST SHIELO~—]| ORNL-DWG 87-4783 CABLE / 1'/2—In. PIPE | | | | 1 v 1 Z n A Ui A h A ~———Cu0 IN STAINLESS STEEL CLOTH BASKET } | GAS | INLET - ‘! . =z Ao e - :?_:_E- ~ TS eveL sar U (| - T T ||l = -d - .- - ONE ¥ -in. HOLE THREE Y5~ in. HOL THREE Y%g-in. HOL THREE Ysp-in. HO ] T F T T T Tt T T o L L L L — S — . 3 Be RODS Y4~in. 0D x 1 SECTION B-B é '/.;-in. LONG ORNL -DWG 69-13105 4 in. BETWEEN WINDOWS S, ES, LES WELDED TAB WITH Y4g-in. HOLE SOLID NICKEL SPACER Lot 3/ -in, OD x Vap~in, WALL NICKEL TUBING T L WINDOWS ———— | A A A —%g-in, OD X 3-in, TALL GRAPHITE CYLINDER | o— WELD ™ A a— 3, -in.~0D Ni TUBE SHOULDER FOR GRAPHITE CYLINDER TO SIT ON—1 -7 Ll N M N N N ' N N N N N M N N N N N » N N N N N N N W h i i — Ya-in.~DIAM PERFORATIONS U U U b 9: :O;-»p A?OUT EVERY Y4 in, [T og_l w | ‘L et N P’ Fig. 18.5d Surface Tension Capsule 316 the atmosphere via the stack., Also the vent valves from the gas operators were kept closed except during openings and closures of‘the access door. The neoprene seal at the access door continued to deteriorate., 1In April of 1967, the increased buffer gas leakage across the inner seal was countered by lowering the buffer pressure requirement in the safety inter- lock system from 50 to hO psia. In July 1967, the safety switch setting was lowered to 35 psia and the helium flow rate to the buffer seal was also increased so that a satisfactory buffer pressure could be maintained. During the intensive sampling of the lest run, the access port door fre- quently required repositioning with the manipulators before the second set of clamps were actuated to lock the door and obtain an adequate seal. The cause of thé detefioration of this seal has not been determined. The radiation level of some‘of the samples taken was in excess of 1000 R/hr at 3 inches and during boot replacements when the sampler was empty, the radiation level inside Area 3A was approximately 100 R/hr. Although some long-term deterioration may be attributed to radiation damage, more likely the reduction in buffer pressure during the last two runs can be attributed to the fact that the left center Knu-vise clamp was loose and thus ineffec- tive in its closed position. The flexible containment membranes (boots) on the manipulator contin- ued to be a chronic source of trouble. Generally, when a leak developed, the sampling cycle would be completed before repairs were made; however, during the remainder of the sampling cycle, containment was abetted by throttling the boot evacuation valve to assure that any leakage would be ‘into the plenum between the boots and through the vacuum pump to the stack rather than to the work ared. After full power was achieved, replacement of the boots became somewhat more involved because the manipulator assembly was quite contaminated. During & boot replacement in early 1967, the radi- ation level of the manipfilator arm was 300 R/hr at 3 in. It was later de- contaminated and saved for possible future use. Also while sampling during early full-power operations, the radiation level at the operating area increased to approximately 30 mR/hr. The radi- ation level was reduced by an order of magnitude by changing the sampling procedure so that a purge of helium from Area 1C to the pump bowl was 317 maintained when the operational and maintenance valves were open. Also on - retrieval into Area lC,rthersamples'were purged for 1-1/2 hr to the off-gas - system before further removel operstions were continued. - On two occasions during the routine transfer of capsules between Areas 1C and 3A, a capsule was inadvertently dropped and rolled out of - sight and out of manipuletor'range in Area 3A. These were later found and recovered by the use of s mirror and a grappling fork which were lowered through the removal valve., ' During the latter part of Run 19, routine sampling weas suspended for approximately four days and the sampler was adapted;to accommodate a col- limator and & germanium crystal detector atop the sampler where the carrier cask is;normslly positibned.j_A plastic plug in the removal area permitted the unobstructed view into Area 3A where samples retrieved from the fuel pump were positioned. Gamma4rsy spectrometry data were then collected on short-lived fuel fission products within 50 minutes of their removal from the circulating salt stream. l8.l.h‘ Discussion and'Conclusions The main operational problems encountered with the sampler-enricher resulted from 1) manipulstor boot failure, 2) lodging of the capsule or _latch at the sampler tube entrance, operational valve, or maintenance valve with subsequent cable tangling, and 3) buffer seal lesks at the operation ~and removal valves and at the capsule access door, | The cause of some of the boot failures were not determined because the leaks were small and the actlons which produced the ruptures wvere not elways coincident with the dlscovery of the leak. Also the boots were not: examined in deta11~because,of the-high radiation level of the boots and | manipulator_assembly, Itpis;believed that the boot failures which could - not.beiattributed to any"spegifie'action occurré@'during'the capsule load- 'ing operation where the manipulator is retracted and pushed downward and | to the left in order to load the sample into the liner of the transport "“tube bottom. At times (almosfirlnvariablyW1th the long capsules), the sampler operator had to resort to pulling the manipulator with both hands in order to retract it.fsrhenough to insert the capsule and latch pin into the liner. Had the capsule loading station been mounted to the left of 318 the manipulator centerline, it could have been located closer to Area 1C and thus require less manipulator travel. Another possible alternative would be to fEb:icete the boots with fever convolutions; however, the con- volution fiidth would have to be increased to maintain the same extended lengfih, The retracted thickness of the boots would then be decreased per- :mitting the manipulator to be pulled back farther than previofiSlyo - Why and where the capsule or latch occasionally lodged in the sampler tube during cepsule insertion has not been satisfactorily answered. How- ever, the use of the proximity switch minimized the possibility of serious | cable'tangling problems, Perhaps two additional detection devices mounted &bove and below the operational valve would have determined the exact - trouble spot. , The buffer seal leak at the access port door started to increase in January 1967 and continued to increase slowly for the remainder of MSRE operations. Perhaps some ‘of this dete;ioration can be attributed to radi- ation damage to the neoprene geskets. The radiation lefiel in Area 3A was 100~200 R/hr during intensive sampling. Also, the accees door was salways in the closed position except during capsule transfer ‘therefore the neo- 'prene was almost always compressed which over the years has probably caused it to yield to a thinner protrusion. In addition, the stellite plates on the access door appear to be gouged slightly by the stellite;tipped Knu-vise clamps. Also the left center_clamp'wae_discovered to be loose in its closed position during the last run. Therefore, it is believed that ordinary wear- and-tear of these components was the primary cause of seal leakage. The beating that the door took could possibly have been iesSened by a flow re- strictor to the clamp actuating cylinders. ' The top seals of the buffered velves developed legks whereas the bot- tom seals apparently suffered only slight deterioration which indicates that particulate matter which occasionelly drop on the valves was the cause of seal deterioration. Consequently the volumes above these valves'showed a gradual incresase in pressure and required constant venting. The least desireble of these leaks was that at the removal valve. On removal of a sample, the transport tube and removel tool which are withdrawn from Area 3A are locked in position asbove the removal valve and the removal” valve is then closed. If an apprecisble delasy is encountered between the N\ g P b i RS L AT prgge ™ W pLLats 7 et e id e L ket et Tl RS SR G TR L s R s e pmmmn e mmm nAsmmmm o 319 time that the valve is closedfand‘the time that the transport tube is with- - drawn from the removsl seaififito'the'shipping cask, appreciable pressure (due to seal leakage) can build;up,intthe removal area and blow contami- nation out.of the removal area to the top of the sampler and shipping cask vhen the transport tube clears the two O-rings of the removal seal. Swab- bing the removal area with wet wipes on a monthly basis greatly reduced the contamination problem encountered in this area. During intensive sam- pling, this area should be cleaned more frequently The specifications for the electrical insulation on the lead wires near the penetrations at Areas lC‘and 3A should be rev1ewed. A failure at the connector pins occurred in May 1966 resulting in open circuits in the insert and withdraw modesiand in the upper 1limit 1light. During in- - spection of the 1C assefibly which was removed in August 1967, the insula- “-tion on the lead wire was found to be very brittle and most of it was un- av01dably stripped durlng removal of the drlve unlt In November 1967 an electrical fault occurred in the same circuits as those in May 1966 except the failure occurred in theewires between the two containment boxes. It is not clear whether these failures-can be attributed to radiation damage to the'insuletion or possibly to excess temperature caused by resistance heating in the lead wire aue;to blocked rotor current (0.3 amps) in com- binstion with the heat supplied by the illuminator lamp, ~ To increase the precision of uranium isotopic analyses, provisions should be made to clean Area'3A periodically, especially after enrichments are made, During the: enriching procedure, -small quantltles of enriching salt could easily adhere to the ‘manipulator and/or be Jarred 1oose from _ the capsule and fall to the floor of Area 3A where all capsules are laid ’-temporarlly vhile swapplng capsules to and from the latch.and the transport - tube, A very small amount - of enrlched salt adhering to a8 sample submitted 1for isotopic analy51s could dlstort the 1nterpretation of the resulting - anelyses. 320 18.2 Coolant Sampler In prificiple, sampling{df the coolant system was the same as that for the fuel system. However, the induced activity in the;coplant salt was very short-lived and there was no residuasl radioactivity in the'sample; therefore, no shielding was1required afld only single containment was neces- sary during transport. Direct manipulation of the sampie by one hand | through a glove part s1mp1ified sampler construction énd operation. 18.2.1 Description of the Coolent Sampler The coolant salt sampler, shown in Fig. 18.6 consisted essentielly of a dry béx connected to the‘éoolant pump by a transfer tube. Two ball valves isolated the sampler from the pump bowl which has & mist shield and capsule guide similar to that in the fuel pump. The 1-1/2 in. Sched-l0 transfer tube was similer to that used for the fuel system exdept,that no expansion Joint was required because the coolant pump position did not change with system temperature as did the fuel pump. The sampler was pro- ;o vided with lighting and viewing ports and the necessary capsule trensfers gfi; inside the sampler were done with one hand through a glove port. Sliding gas seals at the top of the sample carrier along with a vacuum pump and a helium supply were used to reduce oxygen contaminetion to a minimum during sampling and during sample transport to the analytical laboratory. A check-1list type procedure along with a system of locks and keys for valve manipulation was used to assure proper sequence of operations. A key was used to unlock one valve, which could then be opened. When the valve was opened, it also locked the first key in position and released a second key. Removal of the second key (which must be removed to unlock the neit_valve in sequence) assured that the first valve was locked in the proper position. v 18.2.2 Opereting Experience with the Coolant System Sampler The coolant salt sampler operated reletively trouble-free throughout MSRE operations. Only 81 coolant system samples were teken as opposed to T4 cycles for the fuel system. 321 PHOTO 68262 322 Early in 1966 during one sampling operation, the latch failed to slide freely through the transfer tube. No reason for the difficulty'cculd be found. However, the outside - dlameter of the latch was reduced and no simi- lar trouble was reported Also in the first part of 1966 one pin in an 1nd1cator 11ght circuit shorted out at the penetratlon durlng a sampllng cycle. The receptacle was removed and & nev. one welded in place. | | Durlng the same period, the leak rate from the lower seal of the re- moval valve increased. After the valve was disassembled, cleaned, and re- asseflbled; the seai was satlsfactory, however, after apprcx1matelyge1ght additiona1~semples, phe seal leakage again increased and Wae corrected by replacing the valve-seele. ~Also an extra washer was added to the valve to permit more mechanical pressurerto be_applied to the seals. Thereafter the coolant sampler,operated_trodble-free for the finalfthree years of MSRE operation. 18.2.3 Conclusions and Recommendations From an operational viewpoint, two improvements may be worth con- 51dering. A (1) The glove port cover should be hinged to sw1ng out horlzontally. On several occasions when the clamps were removed, the glove port cover. slipped from the operator s hand and dropped freely to its stopped posi- tion. In doing so, it could have pinched & bundle of nearby insulated electrical wires. 7 ) .7 - (2) The sample carrier should be lengthened approximately two inches to accommodate the longer capsules which were used during“the'last run. - 323 19. CONTROL RODS M. Richardson | -19.1 . Description Three control rods were used in the MSRE core vessel, The rods were not exposed directly to the fuel salt but were contained in three "blind well" thimbles which were located in three of the four channels surrounding ~the vertical centerline,of_theigraphite core. The thimbles were straight within the core veeeel; butdabove'the reactor access flange they were off- set — using two 16-in. radius bends to provide space for the core sample containment standpipe. A roller was located at each bend to reduce fric- ‘tion, The control rod drive mechanisms were vertically positioned around the standpipe. The control rods were fdbricated of flexible metal hose with hollow center cyllndrlcal poison elements (38 per rod) beaded over the lower 60 in. Rod motion was ‘supplied by motor driven 1/k-in. roller chain and sprocket drive units. Included in the reverse locking power train is an overrunning clutch whlch permits powver transm1551on in the in- sert direction only. A magnetlc clutch was used to release each rod from - the power. train gearing and permltted free fall of the rod into the core at A0.h g. Continuous indication of each rod position was prov1ded by two synchro "torque transmitter-receiver palrs, one "flne and one "coarse." - A potenti- ometer provided p051tion ;ndlcatlon for safety interlocks in the reactor £i11 circuits._ The upper and lower limits of rod travel were controlled by four (2 upper and 2 lower) mechanically actuated switches. - The poison elements were cooled by cell air (95% Nz) from the . compon— ‘ ent coolant system.r This entered the upper end of the hollow control rod - and "exhausted radlally into the thlmble from & nozzle at the lower end- of ~the rod. | | | | .Changes 1n pressure drop as thls nozzle moved through a flow restrlc—, tor built 1nto the bottom of the thlmble prov1ded a rod position reference ~ point (flducial 'zero)., Comparlson of this with the rod position as indi- cated by the eynchro rod position instrumentation was used to monitor stretching of the rods and other malfunctions. 324 The rod drive units were identical as,fiere the rods except for a slight variation in lengths of the individusl rods due to the differences of the thimble offsets. This allowed interchange 6f_components for repairs or trouble-shooting. For identification of a rod assembly, it was neces- sary to specify the thimble number (T-1, T-2, or T-3), the rod number (R-1, ‘R-2, R=3, or R-l) and the drive number (V-1, V-2, V-3, or V-k). The rod assembly connected to the servo controller was considered to be the regu- lating rod and the other two rods were the shim rods. The rod assemblies could be used interchangeably as shims or regulating rods by shifting the ~out-of-cell rod drive disconnects. Any one rod was capable of teking the reactor subcritical.2! | The rods and drive assemblles were not designed for in-cell malnte- nance, Repalrs on the drlve units were done after dlsassembling the rod from the drive and removing the drive unit from the reasctor cell by remote maintenance ‘methods. The control rod shock absorbers naed the general principles of a rypi- cal hydraulic shock absorber but differed in that the working "fluld" con- - sisted of 3/32-in,—diameter steel balls. At the end of a scram, the bottom " face of the shock ebsorber cylinder struck perfianentiy-mounted steel blocks which were belted to the hofising° The shock ebsorber plunger, to which the control rod was attached, continued to move downward and was decelerated by the fbrees developed by the”bnffer springs and by the flow of the ateel balls around a knob on the plunger° The atrokenlength of the shock ebsor- bers were adjusted to 3.5 inches for each rod, 19.2 Initial Testing The rods were installed in the reactor vessel in January of 1965. The limit switches were adjusted to eoincide with the rod position indicators on the main console and electrical continuity tests were performed on all drives prior to assembly to the control rods. Satisfactory ambient tem~ perature performance waS“checked(with'the_resnlts given in Teble 19.1. / 325 Table 19.1 inifiial-Tests of Control Rod Assemblies ; | | "Spare‘ Thimble No. | - - P-1 - T2 T-3 - Components Rod No. R-l Re2 R-3 R-4 Drive No. - V=3 V-2 V-1 V-l Motor current withdraw —-ampé 0.6 0.6 2 0.6 - 0.6 Motor current insert — amps 0.59 0.58 0.58 . 0.59 Shock stroke, inches = 2.85 3.6 | 3.7 3.0 Scram time, secs’ 0.835 0.752 0.775 _— Air flow, rod — scfm 3.8 3.9 3.9 e Rod travel speed, in,./sec 0.53 0.53- - 0.53" 0.52 Full rod travel, inches 50,91 50,95 50.95 50.9 Fiducial zero, inches 1.k 1.hb o 1,35 - 19.3- Periodic Testing During reactor operation and especially after power operatlons com- menced it was not possible to run extensive tests on the rods. However, it was necessary to do sufficient testing to assure that the rods would scram if needed. It was also important to know that the rod position indi- cators were functlonlng proPerly as these were used in m&klng reactivity zbalances. To accompllsh this and to aid in antlcipatlng necessary mainte- ~ nance; the following perlodlc tests were made, Co / A. Rod Scrams ——-Scram.tests were | routlnely conducted to determlne ‘rod scram times before & fuel: flll snd before critical operatlon. Fach _rod vas withdrawn to a helght of 50 in. The magnetic clutch current was broken by trlpping the menual scram switch. The rod'would free-fall to the fully-lnserted position. - The 50—1n. drop was electronically timed from the moment of release to the 1ower llmlt switch. The maximum permissible scram time was 1.3 sec/50 in. | | o B, Rod Exercise —-Sincé rod scram tests could not be made during nuclear operation, each rod was exercised daily to demonstrate that they 326 | I would move on command and also to flex the metallic hose. The flexible hoses tended to stiffen with time if allowed to remain in one positién for prolonged periods. Observation by the reactor operator of the rod—positién indicators during the exercise ascertained that the synchros were working properly and that the rods operated freely in the thimbles. C. Fiducial Zero — As described above, the fiducial zero vas a fixed reference position in each of the control rod thimbles which related the actual rod position to the rod position indicated by the position-indicating potentiometers. Fiducial zero tests were routinely made before a fuel fill and before critical operation. 19.4 Operating Experience Except for difficulties with the rod assemblies in thimble T-1, opera- tion was satisfactory from Both an operational an& maintenance standpoint. There were no unscheduled shutdowns because of a control rod failure. The- testing program was adequate to allow advence maintenance planning and work was done in conjunction with maintenance in other areas at the time of a scheduled shutdown. | During over 3,000 rod scrams from 50 inches (see Table 19.2) and num- erous others from different elevations, there was only one time when a rod failed to scram. Except for this and one other period, the scram times were all less than the 1.3 seconds specified in the safety limits and wére ‘usually less than 1 second. The number or amount of rod movement'is_not known. Oncerpositioned, there was little movement of the shim rods. The regulating rod, normally the assémbly in thimble T-1, moved little at steady power until bubbles appeared in the reactor. From that point on the regu- lating rod moved frequently under servo control tb compensate'for the action of bubbles on the reactivity.~ Sevéral rod jog tests were made and it is estimated that the regulating rod moved about 720 jogs per hour at 0.l in, _ per jog for a total of 32,400 jogs. The total movement was about 1h,000 inches. 327 Teble 19,2 Estimated Number of 50-inch Rod Scrams Thimble No. Rod No. Drive No, No. of Scrams 1 Rl V-3 750 T-1 R-1 V-1 150 T-2 . R-2 V-2 930 -3 ~ R-3 V-1 800 -3 Rk V-1 k30 Total 3060 The estimated life of the control rod motors was 10,000 hours at a reactor pover of 10 MW (Ref. ”h9) The condition of all drive motors, lu- brication, W1ring, and gears remalned satisfactory throughout the entire operation. These were 1ast 1nspected in July 1969 et whlch time the reac- tor had accumulated 92,805 Mwhrs. . , The hlstory of the MSRE control rods and drlves are best summarized in Figs. 19.1, 19.2, and 19.3., Control rods R-1 and R=2 remalned in ser- vice the entire operating period in thimbles 1 and 2. cOhtioi rods R-3 and R-l were 1nterchanged in thlmble 3 as shown 'in Fig. 19. 3 fThe component 1 failures of the drives are. tabulated in Table 19.3. | | Heat—up and pre-crltlcal salt circulation commenced 1n January 1965 and during the early perlod of this operation, dlfficulty was encountered with the limit switches. The 1ndications were that the sw1tch operators were sticking or galling randomly on all three rods, Temporary repairs ~were made during the March and Aprll 1965 shutdown when'investigatioh re- vealed that the sllding meflbers of the switch operators had galled causing the faulty operetion, - , The drives were removed from the reactor cell in July 1965 for instal- lation of modified limit switch operators on all the units including the T-1 ®TqUFYL UT SOFIquassy poy 3o Azénsrn °o7AI8S 1°6T 374 .. Maintenance History " FIDUGIAL ZERO (in.) W . 'SCRAM TIME (sac) Installed temperature switch Replaced Irimit‘ switch operator Replaced potentiometer —] 0000 8 & § T - —— 0OLO "v 0080 -1 0080 -1 000°} MAY JUNE JULY AUGUST SEPTEMBER & OCTOBER NOVEMBER DECEMBER Replaced temperature switch Replaced potentiometer w - ATTEWASSY 1—U 'E-A JANUARY - - FEBRUARY MARCH APRIL MAY JUNE . JULY AUGUST 9961 SEPTEMBER OCTOBER NOVEMBER DECEMBER ol a JANUARY FEBRUARY MARCH APRIL MAY JUNE JULY AUGUST [ SEPTEMBER OCTOBER L9681 NOVEMBER DECEMBER F June JANUARY FEBRUARY MARCH APRIL MAY Bo6I JULY AUGUST SEPTEMBER OCTOBER NOVEMBER DECEMBER Fine synchro failed Potentiometer noisy Rod No. 1 used as shim rod Replaced V—3 drive with V—4 Rod No. 1 regulating rod JANUARY. FEBRUARY MARCH APRIL MAY JUNE JULY €961 AUGUST |_SEPTEMBER_ OCTOBER NOVEMBER DECEMBER 8ct 209—€L DMA—INUHO Z-1 STQUTYL U SOJTqUIssy poy jo KI03ISTH @9TAleg Z°6T *91d FIDUCIAL ZERO {in.) SCRAM TIME {sec) o o Maintenance History ' g § § . § g -§‘ el MAY JUNE _ JuLYy AUGUST = = X . w—t 00L0 Installed temperature switch { l ' ] ' I Replaced _Iimit switch pm ‘ . OCTOBER NOVEMBER DECEMBER 1 JANUARY FEBRUARY MARCH APRIL . MAY - JUNE = JULY § AUGUST. ‘ { SEPTEMBER OCTOBER NOVEMBER DECEMBER | JANUARY _ FEBRUARY MARCH . APRIL - < , _ e Replaced fine synchro . ’ » L MAY d ' b JUNE g - : JULY . , ' AUGUST SEPTEMBER OCTOBER NOVEMBER DECEMBER , B B JANUARY ° o _ FEBRUARY MARCH APRIL MAY JUNE JULY AUGUST SEPTEMBER OCTOBER NOVEMBER DECEMBER JANUARY | FEBRUARY MARCH APRIL MAY JUNE . - JuLy e AUGUST . SEPTEMBER ot , OCTOBER , Y NOVEMBER | - C . ) DECEMBER l " A18W3SSY Z-Y 'Z'A"""""' s e e - - £961 Replaced fine synchro 8961 — s Replaced potentiormeter 6ct SEPTEMBER @~ €09—€L OMO—INKO Maintenance History FIDUCIAL ZERO {in.) ~N W SCRAM TIME (sec) o 3 8 0060 Installed temperature switch Replaced limit switch operator Repaired roller in T—3 Repaired R—3 upper hose 8 8 & B bhor ] ) —1 00,0 MAY JUNE [JuLy AUGUST - SEPTEMBER §{ { OCTOBER | NovEMBER DECEMBER Bent limit switch operator R—3 sticking in thimble — Replaced R—3 with spare rod R—4 v-uls—u—— ATENISSY Z-H * | —A —masieits *> JANUARY FEBRUARY jMARCH APRIL 9981 OCTOBER NOVEMBER DECEMBER Bent limit swi_tch operator | AT8W3SSY P-4 'L—A = [June lsuLy JANUARY FEBRUARY MARCH APRIL AY I 196 AUGUST | sErTEMBER OCTOBER NOVEMBER IDECEMBER €-1 PTQWIYL UT SOTTqUossy poy JO LI0ISTH 99TAlaS €' *S1d 50 inch scram time increase | SANUVARY | FEBRUARY MARCH APRIL AY - JUNE JULY AUGUST = 961 | SEPTEMBER OCTOBER NQVEMBER DECEMBER Added 1.5 Ib weight to V—1 0.9 scram time Repaired rod R—-3 No. 3 assembly used as reg rod No. 3 failed to scram on demand " Replaced R—4 with R-3 Replaced potentiometer —Y - » e aal g epa—t— 1 ° b A o - AT8N3SSY £-HI—-A S— JANUARY FEBRUARY MARCH - APRIL AY JUNE: JULY AUGUST = 6961 ¥09-€L OMA-TINYO SEPTEMBER OCTOBER NOVEMBER DECEMBER 0te 331 Table 19.3 Drive Unit Component Failures Drive Unit No. = V-2 V-2 V-3 Vol . e , (Spare Drive) Total - S8ervice Time as Reg., Rod 4 mo., = --- 51 mo. 5 mo, _ Service Time as Shim Rod 56 mo. 60 mo. L mo. w=m Temperature switches T - 1 Limit switch operator 1 1 1 1 b Limit switeh operator . , push rod : -2 - - - 2 1000-R potentiometer - - h - | Fine position synchro L transmitter 0 2 1 | - 3 Coarse position : | synchro transmitter 1 ) - - , - 1 - spare (V-4). The new operators contained case-hardened bearing_surfaces, there were no more difficulties of this type with the switch operators. Examinationrof the control rods during this period revealed that the upper wire sheathed upper hose of the No. 3 rod was badly worn at a point 28 in. below the flange. Exemination of the No. 3 thimble revealed that the upper roller, on whichfthe.coatrol rod operates, did not rotate which caused the severe wear te.fhe rod. The roller was-replaeed and & new hose glnstalled on ‘the No. 3 control rod.._ | | -~ During this period temperature SW1tches were added to the lower drive ~meehanlsm. These small bi-metallic switches were to alarm should the tem- ~ perature within the drive unlt cases exceed 200°F., ‘Without the dovnward _:sweep of gas through the housing, convection of the heat rising from the thimble could cause & rapld temperature increase and resultant demage to ?the drive assembly. After theirod.drlves had been reinstalled and the re- actor heated and filled with salt, the switches were fiested.by turning off 332 the gas flow. Since no alarms oécurred, another method of measuring the drive assembly was devised which, although not precise, was adequate for operation. This was done by shutting off the integral drive motor cooling fan and measuring the resistance of the windings of the fan motor. “A slightly greater than- 10% ‘inerease in re51stiv1ty equaled.V50°F change.: w1nd1ng temperature,‘wfrom the measured resistance and us1ng the temperature 'coefflcient of res1stance of the wire, & curve was plotted relatlng the measured resistance to the approximate motor temperature. This value was roughly related to the drive housing temperature. | ' fh- In early operation, the control-rod servo system had not functloned | properly due to coasting of the regulating rod motor and shim locating mo- tor. Brakes were installed in both of these (see 2h.5). 19.4.1 Rod Assembly in Thimble T-1 ’ Figure 19.1 gives the rod drop times and fiduc1al zero values as well as the maintenance history for the rod assemblies located in th1mb1e T-1. This was first made up of Rod R-1l and drive unit V-3 and was usedeas the regulating rod, except for a four-month period, for the entire reactor op- eration and consequently performed the bulk of the control rod service. Other then replacement of faulty drive components, there was little diffi- culty with this assembly.. - , ' ' . - The .1000-R: potentiometer position indicating potentibmeter'failed 3 times due to an open.resistance coil and 1 time dQue to & wvorn resistance coil. This difficulty had been anticipated as it is a common failure for this type.component in continuous service. L : " One-temperatureksvitch failed due to an electricel ground at tfie wire connections at the switch. The ground was created by bumping the sw1tch during assembly of the drive into the case, .. There was &n eapparent.shift of 1/2-in. in the fiducial zero ‘position in November of 1965. ‘The.change was attributed to slippage of the'chain on the sprocket but examination.and testinglfailed-td reproduce the slip. The 1/2-in. deviation was:recovered involuntarily in July of 1967~whiehi wouldfinddcate-that the exact cause of the shift is not known. The fidu- _eiallzerq position plot in Fig. 19.1: shows the;gradualiincrease,,$1/2 inch, 333 in the length of theAcbntrolfrod'from'the effects of usage. The change in length which occurred in September 1969, is the result of changing the drive unit. | | _ ‘The rod scram times remained at 0.8 *+ .05 seconds from installation until the spring of 1969. In March of 1969, the scram time increased to 1.03 sec for a 50~in, dr0prand5also during this period, there #as a failure of the fine position synchro transmitter and the 1000-R pot. Analysis (see 19.5) of the rod acceleration and velocity during a scram fevealed that . there was an area of drag in the lower 20 in. of the scram. From the sbove evidence, the V=3 drive was féplaoed with the spare drive V-L. With drive V-4 and rod R-1 in thimble T-1 the scram time returned to n0.8 sec until ~ shutdown. A detailed inspection and maintenance program is planned for 3 drive V-3 but is as yet incomplete. . o 19.4.2 Rod Assembly in Thimble T-2 - As shown in Fig.'l9.2,_the assemblyrin thimble T-2 was made up of V-2 drive, R-2 rod. This.assembly was used as a shim rod, except for short test periods, for the entire MSRE operation. Other than the mechanical :failures shown in Fig. 19.2,}there Werélno operating difficulties. The mechanical failureo included two fine poSition synchro failures and one potehtiometer;':Ekémifiation of one of the syhohrosl(the'radiation level of the synchro was NlO mR/hr) revealed the failure to be the wiper arm contact between the rotating shaft and the windings. ' ' The fiducial zero history shown in Fig. 19.2 clearly shows the pro- gressive stretchlng of the rod to be ml/2 in. from the tlme of 1nstallation to the end of operatlons.'i,a o ' ~ 19.4.3 Rod Assembly in Thimble T-3° _ This assembly was. made up of V-1 drive and.R—S rod at the time of in- .3stallation es shown in Fig. 19 3. As 1nd1cated by the scram history shown in the figure, this assembly was ‘the most troublesome of the three MSRE 'icontrol rods. The drive units were removed during the precritical mainte- :nance period in the summer of 1965 to install the modified limit switches described elsewhere. The No,_3rcontrol rod was 1nspected at this time and found to have a torn Sectioh'in'the_uppor hose, 28 in. below'the flange. The damage to the rod was the result of the_failuré'of the upper roller A e g e e 334 in the No. 3 thimble vhich had jammed and would not rotate. The roller and rod were repaired and after reassembly, the rod scram time was <0.8 sec. B S A _ During the low power period of operation,'it was found that after a full seram the lower limit switch would not clear when.the rod was with- drewn. _It was necessary to push the switch actuator by withdrawing the. rod to the upper 1limit to clear the lower limit switch. It would function nofmally'until the rod was again scrammed. The bottom end of the limit. switeh push red.was found to be bent below the lower retaining bushing. The rod had been bent by striking the lower housing flange as the result of & weak push rod-springe Repairs consisted of straightening—the;push-rod and installlng a stronger spring. o o , . Later the control rod commenced to stick in the thimble at a point ¥1-1/2 in. above the fiducial zero position. -The rod could be freed by Jogging and #ould'hang in no other pesition. Directjexemination of the thimble was not possible, but the exterior of the rod was examined and ap- - peared to be in good condition. The difficulty appeared to be.that & sharp projection such as a broken weld at the throat section of the fiducial zero position existed on which the rod could snag. However, when the R-3 rod was replaced with the spare rod R-L, it did not snag. This rod operated within limits but the scram time increased slowly fromr0,85 sec in May 1967, up to 1.26 sec in December 1968. Examination of both the drive and rod revealed do apparent defects. Replacing the V-1 drive with the V-2 drive unit in the T-3 position for test purposes made little change in the scram time. Therefore a l,S-lb #eightwwas ‘added to the shock absorber assembly o the orlglnal V-1 drive unlt and 1t was re- installed. This reduced the scram time to less than 0.9 secof ‘ The reactor was shut down in June 1969 by manuslly scramming all the cohtrolrrods. The assembly in thimble T-3. (R—h and V-1) was at 35.0 in. and did nof scram,59 The rod was inserted 0.2 in. and scrammed again, this " time the rod dropped freely Repeated attempts"towmake the rod stick fol- 1owed thls incident bux vere unsuccessful Fig,w 19.h shows the results -of »scram.tests at 2—1n° 1ncrements taken on June 2 befbre the rod was exam- ined and clearly shows the ares of hlgh drag. 1.8 e 14 1.2 SCRAM TIME (sec) 06 04 - 02 -0 ~ Fig. 19 ih'f_jlncr_émental Rod Drop Times 335 ORNL -DWG 69-89T6A L o JUNE 2, 1969 8 | | e avsust 4,1969 0 10 -2 - 30 40 50 .~ STARTING POSITION (in.) 60 336 To correct this difficulty, Rod R-4 was removed and-replaced.by Rod R~3. This was the rod which had earlier snagged in the thimble. Before reinstalling it, the air exhaust tube (a poSsible‘causéof the snagging) was filed smooth to eliminate all the sharp edges. The defective potenti- ometer was also replaced on the drive {V-1l). The greatly improved perform- ance of this assembly (R-l4 and V-1 in T-3) suggests that the trouble was in the rod (R-4) rather then in the drive. The extreme radioactivity of rod 3 (R-3) made it impractical to examine. Figure 19.h4 also shows the results of drop tests conducted on August k4, after all repairs had been accomplished for all three controi‘rods. This was the condition of the assemblies during the final period of operation. 19.5 Improved Rod Scram Testing It was obvious from the No. 3 rod failure to scram that the method of testing each rod by’a single 50-in. scram was inadequate. The method failed to expose any local areas of high drag. The incremental method of testing, scram time vs staftingnpoéftioh, es shown in Fig. 19.4, showed no region of excessive drag for afiy of the rods. In an effort to provide a quick testing procedure that wohld not reqfiire numerous rod drdps, yet would reveal regions_of.abnorQal drag, a procedure developed for the EGCR was renovated. ‘The:output'frdfi the 1000-ohm position potentiometer was - smplified and transmitted by wire to the main ORNL area, where it was passéd through a filter with a 5-Hz time constant to remove transmission noise, digitized at 2000 samples/sec, and stored on magnetic tape. The data was then sent to the IBM—360 computer for smoothing and analysis by a program especially developed for fhis purpose. This procedure when acti- vated during a rod drop from 50 in. gave about 1800 data points during the drop. From this velocity afld acceleration curves were generated. Data teken before repair of the sticking rod assembly (R-4 and V-1 in T-3) showed a region of near zero acceleration between 26 and 40 in., in good agreement with the incrementsel results shown in Fig. 19.4, After reinstallation, all three essemblies were tested by both incremental drops and the single drop computer analyzed method. Both methods showed reasonsbly constant acceler- ations of 10 ft/sec? or mofe° A typical plot of acceleration vs position is shown in Fig. 19.5.5% POSITION - INCHES Fig. 19.5 Rod Acceleration Curve LEE 338 The signal filtering and data smoothing that are fequired.because of &u;? the noise that is inherent in the position signal genérator and transmis- sion system made it impossible to detect with éertainty a zero acceleration zone 1éss than about six inches long if it is more than about six inches below the scram starting point. | 7 " Revised criterie wereAestabliShed-for the monthly testing of the-rods. This consisted of scrdfiming the_rbds from 40 and 50 inches. These two points were Just &bove the normal operating positions of the regulating and shim rods. 'The:criteria were that the 50-inch scram time must'be less than 1.00 sec, the 4O-in. scram time must Se less than 0.90 sec, and the acceleration curves must show no area whiéh_indicates abnormal drag. Ir an area of high drag was noted, the rod wouid be dropped from a point within this area. This scram time must be no more than 20% longe; than normal for that point. Prior to a fuel salt f£ill, the rods would be scrammed from 24 inches, which is their normal position for a fill. The scram times from this poifit must be less than 0.7 seconds and the acceleration curve shbw no abnormality.so 19.6 Mszintenance Experience - Repairs perfdrmedron the drive units usually’cbnsisted of a simple repiacement of worn components which were readily accessible after the unit had been removed from the reactor ceil. In the event of major difficulties, both a spare drive assembly and a spare control rod were available as re- placement items. | , It was possible to perform direct out-of-cell maintenance to the drives without excess exposure to personnel. The radiation level was about 600 mR/hr at the drive case (the_drive unit was removed from its case during maintenance). The overall level of radiation at the drive mechanism was gbout 200 mR/hr and, as én example, the level at the fine synchro trans- mitter which was mounted on the gear case was about 30 mR/hr. Maintenance on the control rod itself after criticality was very limited. Activity on ‘the upper hose was sbout 400 mR/hr at contact and the section containing the poison elements was 20 R/hr at 18 in. after a two-year storage period. 339 Since the four drive assemblies were identical and the spare parts were common to all the drives, the maintenance problems regarding replacement of faulty components was not complicated. However, investigation as to the - reasons for slow scram times, even when the areas of high drag were located (see 19 5), was by trial and error method. Past experience had shown that certaln areas such as the 7/16-1n.-d1am air tube, on which the control rod slides, could become slightly gmlled or bent. These areas were examined, adjusted, and the unit operated on a test stand from the main console. The finai test was to reassemble the complete assembly in its normal operating position in the reactor cell and proceed with the scram tests as described elsevhere, 19.7 Discussion and Recommendations - The overall operatien-of the rods and drives was quite good. There was only one time when a rod feiled to scfam in the four years of use. At no time were reactor operations terminated specifically for control rod maintenance. _Construction'of_the assemblies was relatively simple since very rapid response was not a requirement. The flexible controls rods were adequate but the possibility of binding, breakage, etc., always existed. It was possible to control the rod position closely, within 0.1 inch, from the "fine" position indicator on the console but the exact physical loca- tion of the poison elements was subject to weaknesses inherent to a semi- confined flexible metal hose of this type. The only known position was at the fiducial zero positlon in the thimble to which all. other p031tions were related. ' It required about 12 heurs:tO'open the reactor cell access to approach ~ the control rod assemblies, the actual removal required about one hour, " Location of the drives in a more readily accessible location is recommended. The manner of attachlng the control rods to the drlves and the drive _units to the thimbles was hindered by (a) poor lighting at the bolt loca- “tion, (b) lack of space to insert more rigid tools to perform the bolting which was done from a distance of about 18 feet, and (c) poor visibility from the tool operator position at the maintenance shield. A modified 340 external, rather than internal, bolted flange attachment between the rod housing and thimble which is clearly visible is recommended. The "fine" position cynchro transmitters and potentiometers, such' as were used in these units should be of & more durable type for extended service, 341 20, FREEZE VALVES ) M. Richardson ,20;1‘ Introduction The fiow of salt in the MSRE drain, fill, and proceés systems was con- trolled by freezing afidrthawing short plugs of salt in flattened sections of 1-1/2-inch pipe, called "freeze valves." This method of control was adopted because of a lack of a proven relisble mechanical valve. ' Although méchafiical.type valves would have had the advantage of faster action and. the ability to modulate flow, the "freeze valve" concept had a good record of satisfactory performance, the freezing and thdwing times were satisfac- torily Short;'and the "off-on" type control did not impose any particular hardship. . . - | Désign and develoPment df the MSRE freeze valves began in 1960.%0 The basic design was'estdblished.and_hgd been tésted by - the time the MSRE con- struction had'begun.‘ Howevef, cbfisiderable effort was expended at thé re- actor site before the valves were ready for routine operation. Successful freeze valve operatibn is tqtallyidependent on the mannér in which the heat and/br coolant gas is applied;and controlled. The\effectfibfflthe heaters, type,of service, local environment, etc., made it necessary to "tune" each of the valves for its spécific location. This effort continued until criticality. 20.2 Description df'the;Design'of the Freeze Valves The valves were located in the system as shown in Fig. 20.1. All twelve valves were.similaf;andVWere made of'l—l/2—in.'Sched~h0 INOR-8 pipe. The valve body, oriflattened séction, was cold-formed to make a 2-in.-long . flat, &2-in.,wide, with a l/é-in.'internal thickness or flow area. A cool- ~ 1ing gas shroud was welded around the valve flat to direct the gas for ~ freezing the salt. The annulus formed between the valve body and shroud was 1/2-in. The 3/l-in. gas inlet and exhaust lines were welded to the shroud. See Fig. 20.2. A pair of thermocouples located at the center of the valve inside thé air shroud, and a pair upstream and downstream Just outside the air shroud - LINE 204 LINE 206 87 FV=204 FV-206 342 TO COOLANT SALT DRAIN TANK TO AND FROM COOLANT SYSTEM COOLANT SYSTEM FILL AND DRAIN LINES ‘FV-104 FFT ~#———— LINE103 ———&— FV-105 ORNL-DWG 73-605 REACTOR LINE 519 ARRTERRRNRNNNNNN FV-103 : FV—106 LINE 110 Y TO CHEMICAL PROCESS FV-110, FV=-111, FV-112 FD—-1 DIFFERENTIAL V-CONTFK'.)LLEI"R FUEL SYSTEM FILL DRAIN AND TRANSFER LINES Fig. 20.1 Location of Freeze Valves : '\rk‘- AN Mg. ' LN ¥ ! \\ ALL THERMOCOUPLES 3 NOT SHOWN, THERMAL ~"INSULATION . COOLING GAS / SHROUD - 1 1 } -, - - THERMOCOUPLES - fl e e e e s : r ! 7. / / M CERAMIC | ) ' |sTAflcE ELECTRIC RES coot'x\‘.iflrms MEATING ELEMENTS COOLING GAS OUTLET -~ THERMOCOU Fig. 20.2 Typical Freeze Valve 1 ORNL-DWG, 64-6899 - E£4€ 344 were connected to the controi modules and to the cooling air‘modfilating &fij controller. A single thermocouple spaced 5 in. from the valve centerlifie, upstream and downstream on the top side of the 1-1/2-in. pipe, supplied information for setting the manual heater controls. The heaters on either side of the-valve were individually controlled. | The thermocouples at the valve body were connected to an Electra Sys- tems modular control network (31x modules per- valve) which operated auto- maplcally at preset temperstures to maintain each valve within its speci- - fied condition. The control network of the critical valves, descfibed be;ow; included an automatic cooling air flow controller which maintained the valve center at a constant temperature. The ultimate contfol, however, remained with the modules. 20.3j'Descri§tiofi’of the Operation of the Freeze Valves Valves were usually operated in one of three steady-state conditions or modes. . , (1) Deep—frozen — the heaters were adjusted to maintain the valve at (i) i | . 400—500°F without cooling air. R (2) Thawed . — The heaters were adjusted to maintain the valve at. n1200°F without cooling air. (3) Frozen — The heaters remained as in the thawed condition but . the cooling gas flow was sutomatically adjusted to | hold the frozen valve in condition for & rapid thaw. Operation of the valve required only a hand switch (or control action) to freeze or thaw the salt. When the valve was switched to "thaw" from the "frozen" condition, the air was cut off.and the heat supplied by the shoul- der headers would melt the salt in the valve body. Similarly, when the vaelve was switched to the "frozen" from the "thawed" condition, the air was turned on and the salt in the_valve body would freeze. Each valve had three coolant gas flow conditions: off, full on (blast), or reduced flow.Khold).V[The flow condition was determined by the tempera- ture control modules. Aifihough the exact module setpoints varied, for the purpose of explanation, the following conditions are used: (See 20.4.4 for FV-103 control.) o . Qfij 345 Module No. Setting = | ~ Function FV-1A1 ~ 850°F+, 800°F+ Shoulder temperature range FV-1A2 ~ TOO°F+ ) Low shoulder temperature FV-2A1 500°F '~ Center temperature -- low alarm FV-2A2 ~ 1300°F+. Center temperature —- high alarm - FV-3A1 | 850°F+, 800°F+ ~ Shoulder temperature range FV-3A2 TOO°F+ Low shoulder temperature Assume that the wvalve was thawed at 1200°F. When it was switched to freeze, the blast air (&EM scfm) came on. When both valve shoulder temperatures decreased to 800°F, the "plast" air was cut off and the "hold" air auto- matically came on and held the valve shoulder temperatures between 800°F and 850°F, If the shoulder temperatures continued downward for any reeson, all air was cut off at TOO°F by modules 1A2 or 3A2, If hold air flow was ‘insufficient and either of the shoulder temperatures exceeded 850°F, the blast air automatically switched on by the action of the 1Al or 3A1 module and reduced the temperature to 800°F. ~ On those valves used most frequently, FV-103, 105, 106, 204, and 206, a temperature controlled dlfferential air flow adjustment was prov1ded for the hold air in addition to the overrldlng on-off module control. These control;ers maintained the valve center at a constant temperature. The normal valve'condition during reactor operation with fuel salt was as follows: (See Fig. 20. l ) (l) FV-103 was frozen but in the event of an emergency drain, it was requlred to. thaw in less than 15 minutes. , | (2) FV-105 and FV—106 were thawed and were required to remain thawed 77 even in the event of a power outage.’ In the event of an emergency fuel .draln, the salt flowed to both fuel drain tanks through FV-105 and FV-106, | A normal draln required free21ng one of ‘these valves prlor to the drain to .dlrect the salt to the selected recelver tank (3) Fv-104, and FV's 107 to 112 were deep—frozen. (4) Fv-20L and FV-206 were frozen during coolant salt c1rculat10n.’ These valves were requlred to thaw rapidly to prevent the coolant salt from freezing in the radiator tubes. 346 ~During flush salt'operation, FV-103 was frozen end FV-th was thawed. A11 other valves were deep-frozen. ' FV's 107 to 112 were used only during shutdown perlods for salt trans- fers and additions. The thaw and freeze time for these valves was-not im- portant' When not in use, they were frozen or deep—frozen dependlng on the operatlon in- -progress., 20.&_ Operating Experience After the inltlal difficulties described below, operation of the ,freeze valves was quite satisfactory. Table 20.1 is a tdbulatlop of the freeze and thaw cycles for each of the valves.s Teble 20.1 Freeze Valve Freeze-Thaw Cycles Freeze Velve No. o No. of Cycles 103 o 21 10k | - L6 105 . | 97 106 T8 107 - : - 36 108 - | L5 - 109 o 53 110 1k 112 o - 3 204k : o 5T 206 - ) The valves which were requlred to melt rapldly, less than 15 mlnutes, were timed whenever the system.was dralned These valves were FV-103, FV- 20h and FV-206. Periodically they were tested under complete power fail- ure condltions to assure that the valves would melt w1thin the maximum time alloved. The normal thew time without power for FV-103 was 9—11 min, for FV-204 and FV-206 the normsl thew time was 12—1k minutes. 347 During early operation, FV-10k, FV-105, and FV-106, which were re- quired to remain thawed (>850°F) for ~30 minutes, were also tested under - the same power failure conditions. At an initial temperature of 1150°F, the shortest period of time required to reach the salt freezing point was 367minutes, with'nOJsalt"flowing through the pipe. _ It required approximately 8 hours to bring a valve out of the deep- frozen condition (LOO—500°F) into the frozen condition. Heat was applied selectiveiy during the heatup-to insure that the salt was melted progres- sively from either end towards the center of the freeze valve. This-was to avoid having molten salt trapped between two frozen plugs and possible danger of pipe rupture. The siphon pots were interlocked so that the pot-- temperature was requlred to be greater than 950°F before the assoc1ated valve could be thawed. 20.4,1 Testing of a Frozen Valve After a valve had been frozen, it Was—pressure tested to assure that a solid, leaktight frozen plug had been obtained.' Early testing revealed that due to the-fuel drain line geometry, it was possible to have insuf- ficient salt remaining in the freeZe.valve sfter a drain to make up a solid plug. ‘ o , Investlgatlon, using a. glass pipe model demonstrated that after the draln line had been blown down through the head of salt in the drain tank, ~ the salt remained in the siphon pot and would not flow back into the valve body to fill the void. After blowing down line 103, a 4 or 5 psi AP exist- ed between the drain tank: and reactor when attempts were: made to equalize rthls AP, the salt was forced out of the pot and past the freeze valve back into the 103 line._rf S R | | | - The method which was- used ‘most successfully was to reduce the reactor system pressure by ml/2 psig: which usually allowed sufficient flqw-back tor - £i11 the valve and make a good seal A similar. procedure worked satisfac- torlly for the transfer freeze valves., ' 20. 4.2 Control Modules ' j | During the initial checkout of freeze valve control c1rcu1ts and gub- 'sequent operatlons and tests, problems were encountered 1n the operatlon of the modules which make up the Electra Systems Corporation alarm —} 348 monitor system. The setpoints drifted off the pre-set temperatfirettrip value, double trip points occurred, and failures to respond to alarm'sig- nals were encountered. | ' | Failures and malfunctions ef-the modules were traced to.inferior qual- ityfiComponents and corrosion or oxidation at the printed circuit board con- tacts. These faults were mostly eliminated by replacement of faulty com- ponents, gold-plating all module contacts, and minor circuit changes. The setpoints for the T2 modules were checked at least once per year there-: after, In general, the modules remained within 15% of the preset values - after theuipitial difficulties were :resolved. - 20.4.3° "Heat Control to Freeze Valves ~ As installed, the heaters on either side of the freeze sections were on & common control. Also the heat supplied to the lower section of the siphdnipots adjacent to the freeze'valve heaters was marginal due to the manner of installation.. Therefore, a balanced temperature gradient on ei- ther side of the freeze plug was not obtained. Since the setpoints for: the shoulder temperature control modules were adjusted for a balanced heat ,distrlbution, it was difficult to maintein the valves within the module limits. Separate heater controls were added which permitted separate con- .trol of the heat to each side of the freeze plug: The heater. wattage of .the shoulder heaters was increased from 15 W/in2 to 30 W/in?. These added - features relieved most of the gbove difficulties. ~ -The proximity of FV-105 to FV-106 (35 inches apart) closely related the temperature effects of one valve to the other beyond the limits of good .eontrol, This could only be corrected by a change in the piping. Since this was not practical, the problem existed throughout the reactor opera- _tien. This was especially true during a salt fill from FD-1 or FD-2 since ~ the hot salt in line 103 was 19-3/k in. from FV-106 or 15-1/L in. from - _FV-105. ' : 20.k, h"Comments'on Operation of Individual Freeze Valves - FV-103 ——-Slnce this was the main fuel salt drain freeze.valve, it was required.to thaw in less than 15 minutes. A coolant gas flow rate of 75 scfm was .available for freezing. Heat was supplied to the valve from the aflblent temperature within the reactor furnace and therefore no valve heaters were required. Operation of this valve was complicated by & dual 349 set of operating conditions: (1) Immediately following a reactor fill, the line-103 on the downstream side of the valve remained full of salt for a period of several hoursj (2) Line-103 was then emptied of salt leaving no -salt on the:drain tank side.’ Figure 20.3 shows. the temperature distribug _tlon for both condltlons. In order to meet the fast - thaw requirement and still meet both above conditions, the module:control_p01nts were set to function as listed below: FV-103-1A1 -- Blast air on at_lOlO°F+ no hystereszs — (shoulder temperature. ) FV-103-1A2 —- Low. temperature air cut off — U65°F+ — (shoulder temperature) FV-103-2A2 - High alarm 600°F+ "2 of 3" alarm matrix (center ' temperature) , FV-103-3A1 -- Blast air on at T65°F4 - T19°F+, 50°F hysterisis, (shoulder temperature) FVf103-3A2 -~ Low temperature air cut off — 515°F¢, (shoulder temperature). The "hold air" dlfferentlal air flow controller was normally set to ~hold the center temperature (TE-FV-103-2A1) at 375 to LOOCF., The temperature adjustments required for the proper valve operation 'resulted in only one shoulder control module being effective in each con- dition of the valve shown in the figure. During the initial freeze cycle, the TS-FV-103-3Al1 was the principal control and after the 11ne-103 was drained of salt, the TS—FV—103-1A1 was-the key temperature control since there was no salt below TS—FV—103-3A1. During a thaw cycle, the 1A1 tem- ..... on that’ side of the valve. The 3Al temperature responded very rapidly _fs1nce the source of heat was through the frozen plug and- the amblent tem- perature. Selectlon of these temperature alarm p01nts was carefully made .. after many tests. Once the valve conditlons ‘were establlshed FV-103 operated well ex- :'cept for one unscheduled drain which occurred on April 15, 1969. This was ~caused by an. Upward 50°F drift of the setp01nt of the FV-103-3A1 module from T50°F to 810°F plus an administrative change of the FV-103-1Al set- point from 1010°F to 1050°F, . ORNL-DWG 73-806 Fv103)\ (FV1C 2A1 J\ 2A2 )Fv103\/Fvi A\ 1A1 M\ 1A2 FV103Y FV103 T - ‘ 3A1 N\ 3a2 A\ g , - ’ 1 | - . .~ REACTOR ; TEMPERATURE (°F) 1200 g 8 8 § g O.NORMAL FREEZE, NO SALT IN LINE . 103 ON DRAIN TANK SIDE — 9—11 min MELT ® VALVE F.ROZEN, SALT BOTH SIDES OF FREEZE PLUG . B-1 A1B B 28 1B j | R27 — 30 TE NUMBERS. | | | " © Flg. 20.3 "Tefiperature Distribution at FV-103 0st 351 The upward drift had occurred prior to the previous freezing operation and a short plug (<2 in.),of salt was formed in the valve. The - controls and freeze test indicatedea good freeze., At the time of the incident, the salt temperature was not at_1210°F for which the controls were originally adjusted, but at 1110°F with the cooling gas flow controlling the valve center temperature at 420°F, Figure 20,4 is a graphicflhistory of the incident and it can be seen that, with the FV-lOS-lAl-adjfisted to 1050°F, there was little if any warn- ing to the operator prior to the melt. The valve normally melted when TE~FV-103-1A1 was sbout 1030°F—10k0°F when the fuel salt was at 121Q9F at the reactor outlet. The lAlefiodule'administrative'change to 1050°F-was based on an untested theory that the valve should be adjusted on the re- actor inlet salt'temperature'rather than the outlet. As long as the freeze valve had a normal 2-in, freeze plug, the cooling air could be adjusted to maintain the valve within the normal operating range. In this 1nstance, Vdue'to short plug, the_valvermelted.as-a result of changing the cooling air flow which initiated a heating cyoiéflbeyond the range of the cooling air to control. The air flow #as'routinely.adjusted_50?F opwefdewhen the ffiel salt was at a redueed tempefefuie butethe combination of raising the lAl alarm point, the short plug resultlng from the 3Al upward drift before the previous freeze, and raising the cooling setpoint resulted in the fuel draln.” FV-th — This valve operated without dlfflculty since normally it .was deep-frozen or thawed., It was suff1c1ent1y spaced away from.the line- - 103 so as to be unaffected by a fuel drein. See Fig. 20.1. 'FV's 105 and 106 — FV-105 and FV-106 were difficult to control during e'alfill'or drain operation because of the proximity of the valves to each ‘other and to the 103 line which was common to both freeze valves. See Fig. 20.1. Heat conduction from the hot salt passing through line-103 strongly affected that valve which was to remsin frozen. In an emergency drain situatlon, both" valves were thawed and the fuel would drain to both tanks. A scheduled drain requlred that one of these valves be frozen to | direct the salt to the selected‘drain-tank only., In several instances during a scheduled reactor drain ihe frozen valve thawed whieh resulted TEMPERATURE (OF) 352 ORNL~-DWG 73-607 SALT IN REACTOR VESSEL AT 1?10°F 1100]— TE FV 103—1A1 - 1000}= 800]— 800f— J00l— : | © SIMULTANEOUS ALARM AND THAW—-SALT FLOWING THROUGH _ 1/2in. EMERGENCY DRAIN TUBE 600k ¢y - | 103-3A1 - VALVE CENTER TEMPERATURE RISING, COOLANT GAS ? FLOW INCREASE NOT FAST ENOUGH TO CONTAIN HEAT—UP L i , Lo ER Sttt - 103-2A1 RESET CONTROL POINT TIME Fig. 20.4 Teinperatures During Unplanned Thaw of FV-103 353 in an added salt transfer oPeration to put all the salt into'the proper tank., The condition was most troublesome during a reactor fill operation, ia time—consumlng functlon at best, when the frozen valve would go into ‘alarm and control interlocks would stop the fill or the freeze-valve thaw with the system partlally fllled and drain the salt to both storage tanks, | By proceeding very carefully during the initial stages of the fill this condition could, and usually_was, overcome by stopping the fill with the 103 line filled to above theftee'and allOwing the frozen valve to approach ~equilibrium. However, the last failure of this sort occurred as late as April 11, 1969. 7 _ - | ' A leak in the primary'syStem.became evident after the final-system drain of December 12, 1969 which appears to be in the viecinity of FV-105. This is discussed in Section 5 FV's 107 through 112 ——-There were no difflcultles or unscheduled : _thaws with these transfer valves. FV's 204 end 206 ~— The coolant valves have the same 1nherent fault as FV-105 and FV-106 of being too close to each other. Since the valves were always operated as & pair, either frozen or thawed this did not cause any difficulty. There were no operating difficulties with these valves after criticality. N ' 20;5' Recommendations 1. Freeze valves of thlS type should be so located‘by distance or :separate lines :so as. to avoid any temperature 1nteraction..p", , 2. They should be BO- positioned that the adJacent piping is p051— ":tively full of salt, The valves can. be installed in & variety of ways, ='vertically, inclined etc., to insure there are no v01d spaces in the piping. 3. - Those valves which. are not eritical should be simplified by elimi~ .'F;rnating the automatic modular control feature.- These valves would be less '“expen81ve, less time-consuming to adJust and easier: to control by simple manual air control._’l'f' LR ST ) 4, The. automatic differential cooling gas controller should.be con- sidered for all critical valves, This feature greatly reduced the amount 354 of time required-to.maintainlthervalveS'within'limitsAand_reduced the post-freeze thermal cycling. | , - , 5. The shakedown period for adjustment of the valve temperature con- trols at the reactor was time-consuming for a variety of reasons. Initially the control modules were found to. drift off setpoint and hed double trip points. These difficulties were essentially, but not completely, elimi-- nated by local redesign. There were six control mbdules_per-valve for a totai of seventy-two individual modules to be maintained. Each module con- tains two indicating lemps, "Alarm" and "normalq vhich are wired in series with the control-circuit,} There were frequent failures of these bulbs which in turn affected -the module fimction° ‘Many of.the operators had dif- flculty in understanding the functlons of the modules and were at a loss as to what corrective action was to be taken when an alarm sounded. Al- though the Electra modular system succeeded in controlllng the valves, a simpler system is recommended, . 6. Individual control should be prOV1dedfor each of the heaters which affect the valve temperatures. An excess of hesat is;needed to.overcome heat losses in the valv?e;area° PlaCement'df‘the heaters shouldvbe'given careful consideration to eliminate cold spots in adjacent piping. | T. An,adequate,supply of -cooling gas should be availsble to each valve. A supply of L0—50 scfm for 1-1/2-in. pipe "blast air" cooling would be sufficient for good control. » 8. The modular control setppints were established for normal 1210°F reactor operating temperature at steady state. The critical "fast thaw" valve control setpoints were adjusted for-this temperature as vas the -con- trol point for the automatic cooling_gasflcontrol. Deviation -from ‘the nor- mal temperature for which the control points were adjusted (within 75°F) required sdjustment of the cooling air flow only. This reised or lowered . the valve center temperature to maintain the shoulder temperature control modules within limits. During transient conditions, such as heatup of the system, the valves were severely cycled because of a large'tempersture gradient between two valve shoulders_onrthe seme valve, One shoulder would be sbove the module alarm point and that control would turn on the blast eir. The other shoulder would be below the low temperature alarm point - “and cut.all the ceoling air off. The module controls would allow the O 355 ‘velve to cycle back and forth across this range until the low tempe?ature shoulder would heat enough (QTOOpF) to permit the differentisl controller to control the air flow. In fifaétice'all_air was manually turned off during heatup until the valve shoulder temperatures were above TOO°F. The valve then would remain in the ffozen condition but heat much more rapidly - and- the thermal cycling would be greatly reduced. 356 21. FREEZE FLANGES R. H, Guymon 21.1 ‘Description Mechanicel-type - joints were provided in the S—in, selt piping inside the reactor cell to permit major components to be removed for maintenance. Figure 21.1 is a sectional viéw.of-thé so-called "fréezeiflange" type of joint used. The design was such that the "O" ring was in a relatively cool location. A frozen salt seal protected the ring joint séating surfaces from contact with salt which could cause corrosion when the salt wes ex- posed to moisture. Each freeze flange had six thermocouples; four were located on & U4-1/2-in. radius, 90° apart, one on a T-in. radius and one on a 10.2-in." redius. 21.2 Operstion It was never necessary to.remove a major component and therefore one important function of the fieezé'flangés was never tested at the MSRE, Earlier tests on a prototypé flangé indicated that making and breaking these could be done remotely without appreciable difficulty. The leak rates of the five freeze flanges dfiring eafly operation are given in Teble 21.1. Similar leak rates were observed throughout operation except that one of the freeze flanges (FF-201) 1ocated:near the point where the coolant salt left the heat exchanger lesked more than fihe allowable leak rate;of 1 x 10~3 cm3/sec when'salt-was not ‘in the piping and the cool- ant system had been cooled down., This leak was measured at about 0.2 :'L cm3/sec on August 2, 1966. The flange alweys became acceptably sealed when the system was heated. <. o Although there was some variation between individual flanges and some variation with time, the thermocouples located 4-1/2 in. from the centerline of the pipe were usually between 750 and 950°F. Those on T-in. radii‘were sbout 550 to 650°F and those at 10.2-in. were sbout 450 to 550°F. | The thermocycle history is shown in Table 21.2. When it was decided to extend the operation of the MSRE iOnger.than the original plans, the — - e . Sw— - = - - - - - 357 ORNL-LR-DWG €3248R2 FLANG CLAMP GAP WIDTH BUFFER CONNECTION (SHOWN ROTATED) ' MODIFIED R-68 RING GASKET FROZEN #%e-in. R SALT SEAL " '5-in. SCHED-40 PIPE 358 Table 21.1 Observed Leak Rates of Buffer Gas from Fréeze Flanges LéakqgeéRa%ef(sdd emd/gec) " Y a SystemHot -~ o .- - Drained | System Cold System Cold Freeze - Initial Circuleting After After After Flenge = Heatup Selt .- Runl ~Run.1 Run 3 : K i , 1 x10-3 x 10-3 x 1073 x 10~3 x 1073 100 2,0 . 0.5T 0.7 1.5 2.0 101 1.3 - 0.l 0.25 2,26 1.2 102 . 055 . 0‘3 0.30 . .2033 - 100 200 1.0 0.21 - 0.41 1.48 © 0.3 201 0.6 0.22 - 0.21 - 1.23 - 1.8 freeze flange test lodfi whibh_had cycled & prototype flange for 103 cycles vwas restarted. After cycle'268, examination»using dye-penetrant revealed e crack in the bore in the‘viéinity,of the weld attaching the alignment stub to the face of the flange. The crack appeared about the same after the final thermal cycle, number 5L0. 21.3 Conclusions. . o 1 . i ‘ The freeze flenges performed satisfactorily throughout the reactor operstion. The higher lesk rate on FF-201.yhen‘cold caused some concern but did not require repairs. 359 Table 21.2. Summary,of.Unséheduled Scrams at MSRE with Reactor Critical® | Operating Hours V'Number of Unscheduled Rod Scrams ' . Fuel in ., , ., Human Power b | Ygar Qua::er_ Core . Critical Total Error Failures I&C Opher | 1966 1 6712 62 4 2 0 11 | 2 - 1203 1070 13 2 3 6 2 3 554 - 413 2 0 2 0 0 4 1266 1221 3. 1 1 1 0 1967 1 1861 1852 2 1 0 1 0 2 1254 1186 = 2 11 0 0 3 1318 1292 1 0 1 0o 0 4 2159 2u4. 2 0 1 1 0 1968 1 2048 2045 O 0 0 0 0 2 o 0 0 0 0 0 0 3 88 0 0 0 0 0o 0 4 1000 735 1 1 0 0 0 1969 1 1850 1800 « 2 0 0 0 0 2 1385 1375 3 0 1 0 2 3 1076 1054 . 2 0 0 0o 2 4 1203 1176 0 0 0 0 g Total = 19027 17425 37 8 10 10 9 %rhere 1s no record of ahy'unscheduled scrams during 1965, when fuel ~was in the core for 1062 hr and ‘the reactor was critical (at 1 kW or less) for 230 hr. bestly equipment faults.. For example, five of the last six scrams ~ due to "other" causes occurred when the speed of the variable-frequency " generator- being used temporarily to drive the fuel pump sagged below a : prescribed limit. Co 360 22, CONTAINMENT P. H. Harley 22,1 Description and Criteria " Containment at fhe MSRE was required to be edequate to prevent the escape of multicurieramounts of radibacfivity or dangerous amounts of other materials to the atmosphere. A minimum of two barriers was provided. During operatlon the prlmary barrier vas the piping and equlpment which contained or was connected to the fuel salt system and off-gas sys- tem. Block valves were installed in all the helium supply (cover gas) lines to prevent backup of act1v1ty. Primary containment extended out the 7 off-gas lines and through the charcoal beds. The entire primary system was of all-welded constructlon with leak-detected flanged Joints except at a few less vulnerable locations where sutoclave fittings were used., Es- sentially zero leakage was permissible from the primary system° The secondary containment currounded. the primary containment. It con- sisted malnly of the reactor and drain-tank cells and appendages to them, There are NTOO penetratlons for service lines into the cells., Flgure 22,1 is a simplified diagram showing the various type penetrations and methods of sealing or blocking these. - . Under conditions of the MCA, the secondary containment would be limit- -ed to & pressure of 39 psig by the vapor condensing (see 22.4), The cell temperature would rise to 260°F. Under these conditions the maximum al- lowable 1eekage rate as specified in the MSRE Safety Analysis Report®S was 1% per day of the secondary containment volume. ) ; - During malntenance, the reactor was shut down and the amount of radio- activity availeble for the chances of its release were COnsiderably'reducedv When the primary system was opened, the cells became the pfimary'eontainment with the high-bay area acting as secondary containment. 22.2 Methods Used to Assure Adequate Containment and Results Most of the time the primary and secondary containment was operated well below its design capabilities. Various means were used to assure that containment would be adequate -under the worst conditions. - e e i e et e i R e A «c C ORNL DWG. 672688 . E,‘" s s P et - .. e - o ‘ . + S . oA SIEMBZANE SEALTTT . < . : ’ o P Yol STBHIELD BLOCKS . . . . e — | ] . . - . | . R I\ " . “/ .‘ - T - l. ‘ @ | ’ ag coNTaoy [T | T Sae suery ' : o ‘ N : SAEETY VA. , : AR COMPRESSION ‘ QU\ cw o\sco\.:uthé \\IA"E.& WATER TO STACK = 4, areR S o e o S qob‘rqenr.—c 2% . : : ‘ . . - - . s ; : L ) DAFETY v, . X ' @FE’TY VA. EITTINGS C 'E:Lf:CTgI\_A ‘ NS e WELILT — podew, \ T Cam. & ‘ L -.E\..-...P' -3.9‘5..\(2 -: . . u Ak 4 . : ‘ - L - : | TETAL -TO -cmm\ ' ' TSEaLinG COMPONG. SeAL S %Ar-r.-"v Vi, WEDS . : =eaL o ‘ Ry BEAn \.»"‘:"EALEO CONTAINMENT [ . . \ . 1 PRESSURLED WY " " . WoLOSURE ‘ B ' oo TeoLoE copptl, o FLORRRE (METAL TO GLASS 2 Opota, =3 : . : ' - revy va: S = VA, . - TUERMOCOUPLES SARETY , SATETY WRLOS : ‘ oL EVALOATION 6 Soace ‘ L | _ r ) ruEL ! o1 PUTE C ' conpoufh GASLET Pu:\:u‘-’ ) , Sums n t ( £ TASETY VAL ‘ VELTS - ] o BNTICATED . | TAINTENANCE o S o e R -Tp ENCLOSLEE- . B el IR = ) i ‘ ‘ 80" BUTTELELY VA VUSRBES, SEALS T VAPCD, CONDENSING - SysTEm Fig. 22 1 Schematic of MSRE Secondary Containment Showing Typical Penetration . Seals and Closures ' T19¢ 362 The primary system was pressure-tested annually and the cell air ac- tivity was continuously monitored as an indication that no leak had devel- oped during operation. The secondary system leask rate was checked annually at a positive pressure and was continuously monitored at normal cell pres- sure (—2 psig) during operation. In addition to this, all primary and secondary system block valves were tested ahnually to‘dssure acceptable lesk rates. . 22,2.1 Primary System 'Radiographs; dye checks, and other inspections weré carefully reviewed during’cbnstructibn of the primary system. All flanges had metal "0" rings .which‘were remotely 1eak4detected‘by pressurizing the ring groove to 100 psig with helium, (See_Section'lh«Of'thiS'report.) Although the system was opened 33 timés.for fuel additions, graphite sampling, and maintenance as indicated in Table 22.1, there was no leskage of any flange above the allowable 10-3 cc/min:while the reactor was operating. In 1965 before starting nuclear operatidn, all of the block or check valves were teéted in place or removed and bench-tested. This was part of ‘a; the containment (primary and secondary) check list, Section LE of the MSRE 'Qperating Procedures.zg The specified maximum leak rate on the primary system valves was 1 to 2 ce/min at 20 psig. A number of valves in the | helium cover-gas lines had excessive leskage. This was due to damaged "O" .rings and/or they had foreign particles in them, usually metal chips from machining. After cleaning and installing new “O" rings they had no detect- ablé-leakage. After reinstallation, the lines were pressurizéd-and the fittings-or welds were checked with a helium leak detector. These_tests'were repeated annually. The results are given in Table 22.2 along with results from tests of the secondary block valves which are described later. _ In addition to the gbove, an annual strength test was run on the pri- mary system, This was normally done by pressurizing the fuel system, in- cluding drain tanks, to 60 psig with flush Salt-circulating_at normal oper- ating temperature (1200°F). 'No abnormalities were noted during any of these tests. | 363 - Teble 22.1 Opening of MSRE Primary Containment Flanges - No. of Location ! Times - Primary Resasons - Reactor graphite.sampler. -6 Sémples, core inspection Off-gas line at fuel pump T Line festriptién Off—éas-line-in vent house L ‘Valve,zfilter, particle trap Fuel pump vent line | 2. .fPlugged'valve, modify leak detector Overflow tank vent 2 Plugged valve, access to a lower ' - line : : Drain tank No. 2 vent "1 Fuel eddition , Drain tank No. 2 access -6 Fuel addition, samples, inspection Drain tank No. 1 access 1 Inspection DT to FP equalizer 1 ‘Plugged capillary 'FP upper off-gas line 1 Test oil cateh tank FP rbtary'elemgnt‘ 1 Remote practice, inspection FP level reference 1 Restricted after overfill Total | 33 | Tablei22.2“ Tests of Block Valves " Alloweble - No. of " Number Exceeding L ~Leak Rate =~ Valves Allowable Leak Rate Type.Service At 20 psig ~ Tested 1966 1967 1968 1969 . Heltun lescc/mn. 69 2 3 3 ) Water 5-10 sce/min 23 1 s w - AMrorCell Alr 35 ce/min - 125 5 5 k- 364 There were no leaks in the primary system until after the final drain. At this time a leek was indicated by an increase in cell-air activity. Subsequent investigation showed that this was in the vicinity of one of. the drain freeze valves (FV-105). See Section 5 and Reference 25. 22.,2.2 Secondary Containment. | The criteria for the secondary containment was not as figid\as for the primary containment. Small leeks could be tolerated but it wasrstill necessary to assure that containment was adequate for the worst condltlonso The methods used are described below. | Only one Strength test was made on the reactor and dréin-tank cells, This was in 1962, soon after construction of the cells was‘éomplétedo The ~tops of the cells were closed temporarily by steel membranes and the pene- tration sleeves were temporarily blanked off. The cells were then hydro- statically tested separately at 48 psig (measured at the top of the cells), A leak rate acceptance test wes also made in 1962, The lesk rate was ac- ceptable, however, none of the penetrations had been installed by this ' time.s6 Priqf to powér operation,;the containment block and check valves were tested in place or removed snd bench-tested as described in the containment startup check list (4E of the Operating Procedures).?? The specified maxi- mum leek rate at the test pressure of 20 psig varied according to the lo- cation of the valve but uSually was set at 1 to 2 scc/min for helium cover gas (primary contaifiment),_S to 10 sce/min for other gases and 3 to 5 cé/min for liquids. A considerable number of 1éaks were found and repaired. Leaks which occurred in the water system were mostly in the check valves. . These were the result of foreign particles, apparently washed out -of. the system.and trapped'between the fixed and moving parts of the valves. Two hard-seated valves An the water system had to be replaced with soft- seated valves even though the seats were lapped in an effort to get them to seal. One was a swing check valve in a water 11ne between the surge tank and the condensate tank. The other was & sprlng-loaded hand—operated vent valve on top of the surge tank , The instrument-air block valves with few exéeftions were found to be satisfactory. There were nfimerous leaking tube fitfings in the air lines. 365 ,Several quick disonnects on air lines inside the reactor cell and drain- tank cell were found to be leaking when checked with leak-detector solu- tion. These did not:constitute a leak in secondary containment since each 'g'line has & block valve outside'the cell; but a leak here would affect the cell lesk rate indication when air pressure was on the line to operate the = valve, | | | The butterfly valves in the 30-inch line used for ventilating the cells ~ during meintenance nperationsifiere checked by pressurizing between them. rThe leakage measured by a flowmeter was excessive, and the valves had to ‘be removed from the system to:determine the cause. There was considerable | ‘dirt on the rubber seats, and one was cut. These seats were cleaned and repaired. It was found that the motor drive units would slip on their mounting plates by a small amount, thus causing a slight error in the indi- cated position of the valve. Dowels were installed in the mounting plates to prevent this. Small-leaks:fiefe:also found around the pins'which fasten the butterfly tO’the operating shaft - These leaks were repaired with epoxy resin. The line from the‘thermaieshield rupture discs to the vapor-condensing systems had numerous threaded jJoints that leaked badly when pressurized with nitrogen; 'Each'Joint'Was-bfoken,'the threads were coated with epoxy -resin, and the joint was remade;'.All Joints were leaktight when rechecked with leak-detector solution. o ) ’ The cell pressure vas then 1ncreased in 1ncrements and extensive soap- checking and 1eak-hunt1ng were done._ Alternate top blocks were 1nstalled ‘and the cells were pressurlzed to 1l ps1g to leak~-check all membrane welds - with leak—detector solution._ No leaks were found in the welds.. With the cells at 1 p31g, ell penetrations pipe 301nts tube fittlngs, and mineral- j'iinsula.ted (MI) electrical cable sealsrsubjected to this pressure were [checked'W1th leak-detector solution. Numerous leaks were found in tube fittings and MI cable seals, Manzflof the leaks were stopped by simply tightening the threaded. parts of the seal However, the lesk rate vas still about h500 ft3/day, indicating some major lesk that had not been found. 366 All shield blocks were 1nstalled, the cells were pressurlzed to 5 psig, \fi; end leek-hunting. continued. Three large leaks were locsted. One was through the sleeve that surrounds the fuel off-gas line from the reactor cell to a ventilated pit in the vent house- the sleeve was: ‘supposed to have | been welded to the line in the reactor-cell but this had_beenfoverlooked Therefore, the sleeve vas closed atlthe,vent house, where it was accessible. Another-leak wvas in an instrument air line to‘an'in-cell valve (HCV-523), and‘the third was from the vapor-condensing system to the drain-tank steam domes and out to the north electric service area through a.line that was temporarily cpen. _All_penetrations,.tube fittings, and MI .cable seals were again checked with leek-detector solution. Many MI cable seals which_had' . not leeked at 1 psig were found to be 1eaking, and some .of those which-had been tightened and sealed at 1 ‘psig now leaked Agein, many.of the-seals were tightened. As & result, several of the gland nuts split. and soldering was required. All large leaks vere sealed or greatly reduced, and many of . the small leaks were stopped, | : Leak—huntlng and repairs were continued at 10, 20, and 30 psig.j The _ MI-cable seals as & group accounted for & large percentage of the remaining _‘fiJ lecsks. To stop small-leaks vhich may not have been located, all MI-ceble seals were coatedjuith epoxy resin at the seals outside of the cells. This . was done with the cell pressure-at —l.5 psig so thet the epoxy would be- drawn into the seals. Teflon tape, used extensively on MI-cable seals, ~ other threaded pipe, end tube fittings, did not perform satisfactorily in providlng a gas seal. _ | ( t l | Durlng the 30—p31g test large lesaks were located in both component- coolant-pump-dome flanges using leak-detector solution. (One_of them was audible.) The inlet and;outlet:valves to these domes were thenclcsed,and_ uthe domes were opened for work on the gaskets. The,broad,_flat,ffiiton— rubber gaskets were found to be undamaged., The leakage'waS"attributed to inadequate loading pressure on the ‘broad gaskets, and after they were nar- rowed, no further leeks were cbserved., L L G Leak-rate dats was then taken at 30 20, 10 .and -2 psig with the o component-coolant-pump domes velved back into the system. These are plot- ted in Flg, 22.2 along with curves which relate the ellowable leaksage at various pressures to the allowable leskage at: 39 psig (MCA pressure) for £=J P A < x‘ ‘‘ ¥ -’f.‘ ._w.“._‘ “, TRl 367 ORNL-DWG €66-4753 500 400 300 200 100 CONTAINMENT LEAKAGE RATE (scfd) ASSUMED LINEAR ORIFICE FLOW TURBULENT FLOW -100 e ® O O®O 200 20 o o . CONTAINMENT PRESSURE (psig) LAMINAR FLOW 7 EXPERIMENTAL DATA RELATIONSHIP 20 ~ 30 - 40 - 'F'ig'.”22;'2",'_?_8'é¢6i1&é;ty .' Containm'ent Leak "Ra'iiés' 368 verious flow regimes. Also shown is the highly conservative linear re=- lationship which has no physical basis. The allcwable leek rates based on the orifice flow curve (the most conservetive realistic curve) elong with the measured leak rates during these first tests in 1965 are given in Table 22.3, Teble 22.3 Cell Leek Rates Test: Alloweble . = 7' 0Bserved Lesk Rate Pressure Leak Rate : , “(sefd) ' ~ (psig) (sc-fd)- 1965 1966 1967 1968 1969 - ‘ 30 360 130 | . . 20 290 125 - 35 58 150 * . .. 10 o195 Lo 65 | ¥% ¥ : ~p -5 —20 to =50 =20 to =65 <10 to -3 -15 to —20 —23 to =30 % The instrument air block valves were closed during these tests. % o The leak rates at —2 psig are those calculated for the periods fol- lowing the pressure tests. The instrument air block valves were open during these periods, | Subsequent ennual checks of the secondary conteinment consisted of: (1) testlng and repalrlng ell block valves, (2). checKing the lesk rate at :some elevated pressure (usually 20 psig), and (3) checking the lesk rate "while operating with the cells at —2 psig. - The number of valves found 1e&king durlng_these tests are shown in Table 22,2, The measured 1e§k rates ere given in Table 22.3. o . During oPeratlon of the reactor, the cells were maintained at -2 psig an@ tpe leak rate was monitored somewhat continuously. On three occasions, the reactor was sfiut down due to indicated high cell leak ratés. Further 369 investigation showed that leafiage would not have been’ excessive during an accident on any of these occasions. These are described below. In the latter part of May 1966, & cell lesk rate of 100 scf/day at —2 psig was calculated. The reactor was shut down and subsequent investi- gatlon, including a 20-psig pressure test, disclosed a hlgh leakage rate through the thermocouple sheaths from the pressurized headers into the cells. Since no significant leakage was detected from tHe thermocouple headers outside the reactor cell, a flowmeter was installed in the.nitrogen supply line. This flow wasfincluded in the lesk-rate calculations. Prior to this time the headers had been.maintained between 5 and 50 psig;by periodic pres- snrization. To decrease the pnrge to a minimum, the procedures were changed to maintain the pressure at 5 psig. A containment block valve was installed in the supply -line. | , In November 1966, the indicated call leak rate increased to 300 scfd ‘at =2 psig. . Therefore, the reactor was drained on November 20 Leaks were found in two sair supply lines and one vent line used for in-cell alr—operated valves, One of the supply llnes was capped 51nce 1t supplied & valve which was only operated perlodlcally. A rotameter was installed in the other supply line and the vent line was connected to the cell After measuring a cell leak rate of 45 scf/day durlng a 10-psig pres- sure test,,Run lorwas started. Early in thy run the new rotameter in the air line indicated that the ln4cell leak had increased' Rotameters were 1also 1nstalled on three more alr 11nes vhich were found to be leaking in qhe cell., Later, the known leakage increased to 3500 scfd. Although the lcalculated cell leak rate appeared to remain at about 50 scfd Run 10 was termlnated in January 1967 because the p0551ble error in the measured purge _rates in the leaklng instrument a1r lines exceeded the permlssible leak .rate. During the shutdown, the air—llne leaks were traced to qulck dlscon- _f nects in which neoprene seals had become embrittled. Out of 18 disconnects 'W1th elastcmer seals; elght dlsconnects, all near the center of the reactor cell ‘were 1eaking. Seventeen of these were replaced W1th special adaptors. sealed at one end by an. alunlnum gasket and at the other by a standard metal-tubing compression flttlng. (One disconnect was not replaced be~ ‘cause it was on a line that is always at cell pressure.) No similar 370 difficulties were encountered. After sealing the cell, the lesk rate wa's 50 Scfdo - . o ’ o N 22,3 Discussion of Cell Lesk Rate Determinations - There were at least three ways of determining the cell lesk rate. - These were:- (1) a material balance plus compensation for pressure and tem- perature using changes of the differential pressure'betfieen the cell atmos- phere and an in-cell reference volume; (2) a materisl balance. plus compen- sation for absolute temperature-changes in the cells, and'(3) an-oxygen balance, 7 o The first method was used for calculating all official leak rates at the MSRE, A discussion of each method follows., | Method 1 -- Materiel Balance Plus Compensation Based on the Differential Between the Cell Pressure and the Reference Volume Pressure In order to. determlne the cell leak rate in =& reasonable time, very accurate indication of cell pressure changes was necessary. The instrument (iiJ used to determine the pressure changes was essentially a "U" tube manometer ‘(Plgo 22.3) called a "hook gage. Pointed shafts attached to'micrometers | protruded through "o" rlng seals in the bottom of the chamber. . Readings - taken by adjusting the point of the shaft until it was at the 1iqu1drsur- face and then reading the mlcrometer. Thus changes in water 1evel corre- sponding to cell,pressure changes could be accurately measured The or1g1— nal gage had a range of 2 inches of water. A later model'was dbtaaned in whlch the reference chamber could be moved a measured amount whlch gave a range of 12 inches of water., The changes in cell pressure were measured relative to that of an in- cell reference volume. The reference volume was distributed throughout the - reactor cell and drain-tank cell in an attempt -to compensate for cell tem- ,Aperature changes° It was found that very small leaks in the 11nes to the ) reference volume could give large errors in the leak-rate data. Weldlng .these lines ellmlnated this difflcult.y° Approxlmately 5% of the contained volume was located outside of the reactor and drain-tank cells and_had no temperature compensat1on. Thls f } 371 ORNL DWG. 67-2689 TRANSPARENT CHAMBER -( X LIQUID | LEVEL —7 TEMPERATLRE . - | WOOK GAGE- COMPENSATING ' , ¢ REFERENCE VOLUME'j CONTAINED SAFETY ™LOCK VOLLME- L VALVES (nc.ovc, Lgs0) - Fig. 22,3 Schematic of Hook Gage and Piping for Determining Pressure Change - of Contained Volume Relative to the Reference Volume 372 included the 30-inch-diameter reactor cell ventilation duct (1line 930) in the coolant drain tank cell and the component-coolaent-pump domes in.the special equipment room. These areas were sealed off as well as possible to minimize temperature changes. At pofier there was still e_oonsiderable emount of air leskage pest these from the main blowers which caused day- night temperature variations up to 20°F during winter months., The effect of these temperature fluctuations was magnified by the presence of water vapor 1n the cell air. (Since June 1967, there was & small continuous vater leak (<1 gal/day) into the reactor cell. ) This water evaporated in the cells and condensed in the cooler sections of the containment (1ine 930 and the component-coolant—pump gas cooler) This weas periodioally drained from the system. In colder weather more water would condense, | Reactor power also had an effect on the indicated cell leak rate due to cell'temperature:changes. When the .reactor was teken from zero to full pover, gamma_heeting in the thermal shield and-otfier'equipment in the cell caused the average cell temperature to increase 4 to 8°F and indicated a high cell leak rate for approximately 2k hours.v-These changes were not - adequately compensated for by the pressure referenee rolume. After the cells were closed following each'in—oell.maintenance period, the containment was purged with dry N, to remove ox&gen, so-the cell gas started out dry As the water evaporated, the initial cell leak rate was usually high (75-130 scf/day hes been calculated) but gradually'decreased to an equilibrium value in 5 to 7 deys at which time condensate started to form and was drained from the system. During-this initialtperiod until condensate appeared, we relied on results of leak-testing at positive pres- sure before the ¢ell was evacuated and purged. Due to the scatter in the data and the relatively small change in pressure or temperature that repre- sents-e*large“leak'rate,* considerable'time was'required to obtain data from which relisble lesk rate could be determined, To minimize the tem- perature and other effects; initial and final ideta was usually taken at i ) . i . . : _ A change in cell pressure of 0.3 incges of water (0.0l psi) per day or & change in the average temperature of 0. 0h°F will change the indicated leak rate by about 10 .scfd. 373 the same.time of the day-and at the same operating conditions. Calcula- tions of intervals of several ‘days were'more-consistent than shorter Vperiods; Figure 22, h'isiasfilotgef the data used«tordetermine the leskage rate at 30 psig in 1965. - | ' . Typical flow rates for the materlal balences were: 9 scfd to the sump bubblers, 13 scfd purge for the thermocouple header, and Orto 1000 sefd evacuation flow., Since the evacuation flow was normally the largest and was changed_more often, its sighal was sent to the computer which inte- grated it and each shift typed out a value for the total volume evacuated. Method 2 -~ Using the Absolute Cell Pressures and Temperatures This method was not used because the absolute cell pressure indication was not as accurate as the indication of differential pressure by the hook gage and compensatlng for temperature changes using absolute temperatures* did not give as consistent results as those obtained using the reference - volume., ' Method 3 -- Using an Oxygen Balance The - MSRE containment was purged with N, to keep the O3 concentration <5% primarily toleliminete the danger of an explosion if an oil lesk should develop in the fuel-pump lubricating system. This made keeping an oxygen balance on the cell appear as an attractive method of measuring the cell leak rate when the cell was at a negative pressure. The precision to vhich we- could read the Ojp analyzer was only *0,1% ‘which is equivalent to m60 ft3 of air in the cell. With a leak rate nor- ‘mally %20 ft3/day, it became apparent that the only accurate calculation - would be. over very long tlme periods. In comparison to Method 1 (cell lesk rate using the cell pressure change and a.flow balance), the oxygen balance leak rate was consistently,lowrby v15 sef/d&y. In fact, when the leak rate '-caléulated by Method‘l-was'?15.sef/day5 the 05 data indicated a negative lesksage. | | | % , 7 Ten ambient thermocouples were located in the reactor cell, 6 in the drain-tank cell, and 1 in the special equipment room. 374 < ORNL- DWG 65-13149 2.2 - s ,.;\ - | ,§ oLs g s § e @ g’ \ /.\\;/‘"K; g DRAN TANK CELL AVERAGE \\b // A /h“5—~ g i o, / \ T : o ¥ /] \J/ 21.0 ' 89.2 e i 4 . ™ - \ : fi \-'—-o/ St gy, | E 88.4 REACTOR CELL AVERAGE & SPECIAL EQUIPMENT ROOM TEMPERATURE (°F) PRESSURE CHANGE (in, H,0} ? 14 © B 20" 2 24 o2 04 06 08 0 ?2 4 6 s —ee ({27 =65 —tnj g - 1-28-65 “——'———. s - - -!_ Fig. 22.4 Cohtainment‘Conditidns During Leak Rate Test étVSO psig 375 . Errors involved in the pressure method of calculating the leak rate would probabiy be associated'fiith the rotameters; Eutlthis would imply a 25% error in rotameter calibration. All of the rotameters involved were disconnected and bench-callbrated at least twice. They were found to be accurate to well within 5%. The containment cell was nominally at —2 psig, but Just;downstream.of the component-cooling pumps, the pressure was +6 psig. It wes thought that if gas were leaking out here and into the cell at a proportionately higher rate, we might account for the anomaly. - Simultaneous solution of the leak- rate equations for this situation showed that although a solution was mathe- matically possible, the in-leak would have to be at 3% 0, and the out-leak at’ 20% 0y whlch is not practieal. These results caused us to thlnk that somethlng in the cell was chemi- cally consuming some of the oxygen. Removal from the cell air of %2.5 '_ scf/day of pure 0, would account for the leak-rate discrepancy. One.suggested‘mechanism-for chemically removing'the O, was by oil de- composition. If some of the oil from the component-cooling pumps were con- ~tacting hot pipe (say at FV-103 or in control rod thimbles) it would prob- - ably at least partially decompose according to the approximate relation: / Cell air samples were takern to determlne the COz content and they showed '"NO 1% CO». This was at least three times higher than the CO, content in air, but was less than one would expect to see if all the 02 were being consumed'by 0il decomp051t10n.,' ' ' The containment cell and support structure was composed -primarily of "carbon steel so another prime suspected oxygen depletlon mechanlsm.was rust. “'If all the m1331ng o7} were being consumed to form Fey03, it would take gbout 120 g/day of Fe. Slnce there 1s Ebout 5000 £t2 of carbon steel sur- face area in the cell, this would correspond to a corr031on rate of only 0. 5 mils/year (neglectlng the support structure and plplng of component coolant system). 376 This analysis led to the conclusion that enough oxygen was being re- | moved chemically by oil decomposition and rusting to produce the discrepancy that existed between the different methods of cell lesk-rate calculation. 22,4 Vapor Condensing System An accident can.be'conceiied_ip which molten salt and water could simultaneously leak into the reactor or drain-tank cells. The guantity of . steam produeed could be such that the pressure in the cells-wbuld increase -above design pressure. A vspor-condensing sYstem‘was previded to prevent the steam pressure from rising sbove'the'39—psig design pressure'and to retain the non-condensable gsses. A 12-in. line connected the cells to the vapor-condensing system, This line contained two rupture dlSCS in parallel (a’3-in. disc w1th a bursting pressure of 15 psig and & 10-in. disc Wlth a burstlng pressure of 20 p51g). The line from the rupture discs went to the bottom of a 1800-ft3 vertical tank which contained 1200 ft3 of water to condense the steam., The non-condensable gases went from the top of thls “tank to & '3900-ft3 retention tank. 1 " The tanks were pressure-tested‘by7the'vendof"to 45 psig. After being connected to the reactor cell, the system was leak-tested at 30 psig during initial testing of the reactor and drain-tank cells in 1965. During the annual cell leak tests at pressure, the vapor-condensing system was also pressurized. This protected the rupture discs and provided an integrity test of the vapor-c ondenszng systemn, o The water used in the vertlcal tank contalned pota351um n1tr1te—) potessium“bosate‘as a corrosion inhlbltor. Based on annual samples, one - inhibitos addition was made. There was no increasein_the.iron cpncenfira— tion which indicated little or no corrosion. . _ o Inltlally L level prdbes were provided to assure proper water 1evel ,However, after one of the 1ow level probes falled a bubbler—type level 1nstrument was installed for measurlng the level perlodlcally,,' Slnce there has been no unplanned increases in the pressure of the ' reactor and dra;n-tank cells, the vapor-condensing system hss_pop been usedo 377 22.5 Recommendations_ Leak~checking the numerous valves required disconnecting many lines, partlcularly tube fittings and autoclave connections., Disconnecting nu- merous fittings greatly increases the probability that one or more will o lesk when connections are remade. For this reason, an effort should be made in the-design_of the piping to minimize the number of disconnects necessary for leak-cheCking. For example, nitrogen lines could be con- nected as shown in- Flg. 22.5 to facllltate checklng both the safety valve and the line 1n31de the cell. All block valves should be of high quality and soft-seated. Accessi; bility to the items to be checked and the points used in checklng,them should be considered. Flexibility should be_provided:in the lines which must be disconnected. - : ' Lesks from in-cell sir lines cause errors in 1eak-rate-calculations. When disconnects are necessary; the effect of flux on materials on con- struction should be considered. | ‘ MI-cable seals of the type used at the MSRE (brass and stainless steel) should be modified or another-type used.'debstituting ferrules of teflon, graphite—impregnatedrasbestos, or other relatively soft material for the brass ferrules may be sufficient to prevent 1eakage through them, However, radiation damage and sheath temperature must be con51dered Stan— dard plpe-threaded connectlons for stalnless steel to stainless steel joints should be avoided where poss1ble. Teflon tape should not be used on threaded connectlons for a gas seal -nor on small lines where fragments cen enter the line and clog it or pre- vent check valves from seatlng properly. All pipe, tub1ng, and valves should.be cleaned internally prlor to installation. MI-sheathed thermocouples are preferable to the type used. However, __1f cost or other reasons (1. e. seals at the ends) preclude using them, the initial design should provide forfipressurlzing the terminal headers and measuring the leakage. Soldered Joints in the header system are recom- - mended where practical, Gas analyzers used to continuously monitor the cell atmosphere should be located in a warm area to prevent moisture from condensing in them, by i O\ Wall ' //T////// * 7 i X ORNL DWG. 67-2690 CLOwW meTen, SEALED L/ U — / T\ ! - ®mApSS va, —z | | | l - I I : ‘ l 2 [ e e e e CONTROL | VAL.VE" = C e B J {iwston, mn, SuspLy WEADER e SARETY PLOCK VA, N, evi. Fig. 22.5 Schematic of Instrument Airlines with Leak Checking System. 8LE 379 Fluids used in manometérs or similar gaées which are continuously connected to the contained volume such as the hook gage should have a low vapor pressure to avoid a signifiéant loss during operation. The piping to the gage should be pitched to prevent accumulation of_water and should ‘not be connected to the bottom of temperatuie compensating or reference volumes. | ' | | Tempersture compensating volumes should repmesent the entire contain- ment., The 30-in. cell vent line and the component-cooling system enclosure (5% of celi_vnlume) had no reference volume. The temperature reference volume should be distributed throughout the cells. The MSRE used 6-in. vertical pipes on a weightedrvblume basis in the reactor and drain-tank “cells. Smaller'diameter piping with the same total volume might have given - a more representative temperéture compensation., _ ' The reference volume'Systém external to the cells must have a minimum volume and should be located in an area'held at as constant a temperature as fiossible. Sufficient cell temperature thermocouples should be located - in all sections of the cont&inment“to enable calculation of average tem- perature and temperature changes as.accurately as poséible. If a sensitive cell préésure.measuring device similar to the hook gage is used in future reactor containment, it should be designed to meet sec- ondary containment requirements so that block valves are not required. However, if block valves are needed to isolate the instrument, a separate circuit from the other block valves should be used. Pressure readings are required above the preséfirés,axuwhiCh the other block valves are closed. Safety jumpers could be usedftb'keep this eircuit opén'duiing cell~pressure 'testing. 380. 23. BIOLOGICAL SHIELDING AND RADIATION LEVELS T. L. Hudson The crlterlon for the MBRE biological shield design was that the dose rate would not exceed 2.5 mrem/hr during normal operation at any p01nt on thie shield exterior that is located in an unlimited access area. - Since the MSRE had to fit wifhin an existing reactor contaigment_cellrand build- ing, the shield design allowed for addition of shielding as needed to re-. duce radiation.level at localized_hot spots. The shielding calculatiohs are reported in Reference 7. _ ; The reactor vessel was completely surrounded by a water-cooled steel- ‘and-water-filled thermal shield. The thermal shield and fuel circulating loop were located in the reactor cell. The top of the reactor cell had two layers of concrete blocks. An annular.space filled with magnetite sand and water provided the shielding for the sides and bottom. See Fig. 23.1. When the reactor was not in operation, the fuel was draijned to one or both drain tanks which were 1ocated in ‘the draln-tank cell. * Magnetite con- crete walls faced all accessible areas and the top consisted of two layers of concrete blocks. 23.1 Radiation Sfirveys — Approach to Power ExtensiVe'health-physies surveys of the reaetor'aree'were performed during the initial approach to power. These surveys were made to bring to | light shield inadequacies. With the exception of the areas dlscussed in the fblloW1ng paragraphs, the shielding was found to be adequate. 23.1.1 Coolant Drain Tank Cell | At 1 kW the scattéring of fast neutron and gammsa rays from the reactor cell into the coolant drain tank cell through the 30-in. reactor cell ven- tilation line 930 caused radiation readings of 8 mrem/hr gamma and 50 mrem/hr fast neutron near the exit of the line. At 25 kW very high readings were found again at line 930 (70 mrem/hr gamma, 600 mrem/hr fast neutrons and 30 mrem/hr thermal neutrons). A‘wall-of_16 in. of concrete blocks and 6 in. of borated polyethylene was builad adjacent to the 930 line in the coolant drain tank cell. In addition, tfle reactor off-gas line 522, in the - HOLD-DOWN ~ SECTION "xx" w ® a4 m — | I m : o 7 . ~ : 7, \\.\h.......l..lol.n..lufi.flfhfl \\ m \ < — .//).0 , .l/ o oLw&wV\ JJV@ z |||.|||I..I|u‘\| — ] | O \ e\\\ . / 14. ,//%u /.r ..HrL.... [ % off/ \ / o — ¢e \ wo ===l % & =——5cs N ofy 9 t i MO Y = = & s Lt N E 3 A Bl e 1z . oN . \\o _ m_ / : ny//l _ 3N 7 \.\9 : w , '\ = ~? 77 1 e v LS ; y C o=t / ° ny( | {J \_J \\o\e \ MMW _ g ,/ ° 4/ \\X A . m . © /0' ! -\\\l\@\\ue o Illl.llrr. _ nl\\\\\\\ : ANNULUS- Fig. 23.1 Shield Block Arrangement at Top of R_eé,c_tor Cell 382 coolant drain tank cell was found to be giving a high background to the area. At the end of the 25-kW run, the reading was 100 mR/hr at 1 in. from the 3-in. thick lead shield around the 522 line. Later at higher power levels it was determined that part of the rediation was from the drain-tank vent line 561. Therefore this line was shielded with 3 in..of lead. Even though shielding was added inside the coolant drain tank cell, the radiation at the door (500 mR/hr gamma,llso mrem/hr fast neutrons, 75 mrem/hr thermal neutron) and halfway up the acceSSfremp;(22 mR/hr gamma, 3 mrem/hr fast neutrons, 32 mrem/hr thermal neutron (was hlgh during full-power operation. This aree ‘was clearly marked'with radiatlon zone signs at the entrance to the ramp and the door at the bottom of the ramp was locked durlng nuclear - operation to prevent entry into the cell, 23.1.2 North Electric Service Area o When the reactor power was raised to 1 MW in April 1966;'the radiation level in the NorthElectric Service Area (NESA) was found to be high: 20 _mR/hr on the belcony‘and 8000 fiR/hr at the west weii Investigation showed that there was radioactlve gas in the lines through which helium is added to the drainitanks. Two check valves in each line prevented the gas from getting beyond the secondaryucoptainmentrenclosure, but the enclosure, of 1/2-in. steel, provided.little-gemfia shielding. The pressure in the fuel system at that time was controlled by the newly installed pressure control valve with rather coarse,trim, and the pressure fluctuated around the con- trol point (normally § psig) by & ¢2% '»These pressure fiuctuations caused fission product gases to diffuse more rapidly into the drain tanks and back through the 1/h-in. 11nes through the shield into the NESA. The radiation level was lessened by installing a temporary means of supplying an inter- mittent purge to the gas-addition lines to sweep the fission product gases back into the drain tenks. During the June shutdown, a permenent purge system was installed to sfipply a continuous helium purge of TO cc/min to each of the three gaes-addition lines.. This was proved successful by sub;ft sequent full-power operastion in which the general background in the NESA was <1 mR/hr. | ) 30 mR/hr gamma was found on the southwest corner ebout 4 ft above the floor. 383 A 2-ft by h—ftEby 1-in. thickriead sheet was attached to the wall over the ~ hot spot and s, radiation zone was established. Vent House -—-During the initial approach to full power, stacked con- crete blocks were added to,the~floor area of the vent house, over the char- coal beds and between the_ventfhouse and the reactor building to keep dose rates low. Very narrow beamsficoning from cracks were shielded with lead bricks., Even- so, the background radiation level in the vent house was ',NT mR/hr at full power. The vent house was established as & radiation zone area. . o | f | o o Water Room ——-Induced act1v1ty in the treated water rose to an unex- pectedly high level during Run 4. The actlvity proved to be 12, A hr *2x produced in the corrosion 1nhibitor. A survey of possible replacements , for potassium led to the ch01ce of-lithiumg-highly.enriched in the 713 isoQ- tope to minimize tritium production. See Water System (Section 12) of this report for additional- detalls.~~ | Top of Reactor Cellif—-Itgwas’expected that additional shielding would be required directly above the'reactor, where there are cracks (¢1/2 in.) between the'shield'blocks;fffiuring a”fullépower'run:inlJui§'1966,'two very narrow beams-of'y_with;neutronjradiation'were found between cell blocks for the first time.-'These“read'np,to‘lo rR/hr gamma and 60 mrem/hr fast neu- tron. They were properly marked. 23,2 f Radiation Levels During Operation Durlng operation at full power, the rediation levels in all opera- tional areas were acceptable.“ Some narrow beams were noted from time to - time. These were malnly in the vent house. Due to. induced activaty in the i treated water, areas near 1arge equipment such as the surge tank and heat 3fexchanger were treated as radiation zones. Periodic rad1ation surveys have funot disclosed any appreciable changes in radlation levels. Some typical radiation readings inside the shieldlng are glven in ff'Table 23,1, Changes in radiation levels follow1ng a shutdown from power are shown in Figs. 23.2, 23.3, and 23.5L, ' ' Tgble 23.1 Radiation Dose Rates in Various Aress 384 During and Following Full-Power Operation Gamme. Dose Rate (R/hr) . b Reactor' ' - _ Reactor Drained and LOCATION At T MW 10 kW Drained Flushed Reactor Cell 7 x10% 5.4 x103 2.4 x 103_ 2 x 103 Drain Tenk Cell 4.2 x 103 4.2 x 103 2.6 x 10 | Coolant Cell 100 | Fuel-Sampler-Enricherd 1000 500 A Off-gas Samplérd % trer 5-hours operation at 10 kW which followed sustained operation at full power. bImmediately following the\ébove 5-hr operation at 10 kW, CPwo deys after the above fuel drain. dInside the sampler shielding during sampling. _ RADIATION (R/hr gamma) 385 ORNL-DWG 73-608 REDUCED POWER FROM 105 |— 75MWTO 10KW , | O EAST WALL REACTOR CELL — RM 6000—1 5 '® OVER REACTOR VESSEL — RM 6000—2 . 104 108 o ’.-- kDRAIN FUEL SALT TO FD-2 T .0 1 27 -3 45 8 -y | 8 9 10 11. 213 S .: DAYS DECAY ' LT - Fig. 23.2 Gamma Radiation Level in Reactor Cell 386 ORNL-DWG 73-609 105 5 FUEL SALT DRAINED TO FD-2 2 ‘© E E g £ . ] S ~ L < ~h o ~y < [ ~. o fi\- oy, ha 5 2 Yy 103 0 1t 2 3 456 7 89 101112131415 1617 18 19 2021 222324 25 DAYS DECAY Fig. 23.3 Gamma Radiation Level in Drain Tank Cell (Rm 6000-6 between FD-1 and FD-2) v 387 ORNL-DWG 73-610 EDUCED POWER FROM 309 .5 MW TO 10 KW 2 RADIATION‘ (R/hr 3 ft from coolant pump) 3, -DRAIN FUEL SALT 100 9 ngUE TO OPERATORS TRAINING SESSIONS THE POWER VARIED BETWEEN 10 KW AND 100KW | 10~1 , — _ ; : | o 1+ 2 3 4 5 6 7 8 & 10 1M 12 13 | DAYS DECAY Fig., 23.4 Ganuna-Radiation Level in Coolant Cell (Rm-6010) 388 23.3 Conclusions ~ The biological shielding wasradéquate_as désigned éxcept for the few locelized areas discussed previously. Periodic radiation surveys have not indicated any shift or deterioration in any of the shielding. 389 24, INSTRUMENTATION. dR. H.. Guymon 24,1 Introduction It is not within the-sccpe of this report to ccverrin detail the per- - formances of all instrumentation;WEAn attempt has been made to review it from an operational viewpoint end/report significant items. Information on instrumentation which'nesfbeen given in previous sections on the perfor- ~mance of the entire plent'cr'individual systems and components will not be , repeated here. The performénce of the on-site computer is covered in ‘Reference 58, 2h 2 Descrlptlon Most of the 'MSRE 1nstrumentetion wes of conventional type found in ;other reactors or chemlcal plants. -Empha51s was placed on assuring ade- quate containment and reduc1ng radiation damage. All c1rcu1try was de- signed to fail safe. A very brief descrlptlon follows. _2h 2.1 Nuclear Instruments - The primary elements of all of the nuclear instruments were 1ocated ~in a 36-in.-dia. water-filled thimble.which extended from the high bay ~ through the reactor cell to the vicinity of the reactor vessel. (See Fig. 24.1.) e | o Three uncompensated ion chambers and their &SSOCl&ted fast-trlp com— parators provided safety instrumentatlon for scremming the control rods. ".(High reactor outlet tempereture also caused e rod scram.) These .were used ~ in a two~out-of-three configuration.s_:' Two f1ss1on ch&flbers were provided These chembers had ‘automatic po- :—731t10n1ng dev1ces.r Thelr count rate and “the effect of their position were 'm_ffactored into a 31gnal whlch was recorded on a 10—decade 1oger1thm1c power "',fdrecorder.: They were used to prov1de rod inhibit, rod reverse, and other mfiijcontrol interlocks., 390 ORNL-OWG 64-62% I 241t 2in. B - ' & FLANGE FACE NEUTRON INSTRUMENT JUNCTION BOX : HIGH BAY AREA 4«* 50’ - : FLANGE WATER-SAND ANNULUS et - T, ; FINISHED FLOOR v B e ; EL. 8521t Oin. \ 1 EL.851# 8in. : WATER LEVEL W 849 ft Bin, i @Q' T ‘o > WATER LEVEL INSTRUMENT = CHAMBER GUIDE =+ TUBE {TYPICAL) RTfiflcTOR R - ——¢ END OF 48-in. 00 SLEEVE ERMAL ~ EL. 841N 13 in. SHIELD N % 48-in. 00 OUTER SLEEVE ASSEMBLY (WITH EXPANSION JOINT) ~~EXP JOINT - _ 4 CONTAINMENT PENETRATION EL. 834t 82 in. REACTOR VESSEL - C EL.830ft 3in. \ AN \msuunou REACTOR CONTAINMENT Fig., 24.1 Elevation View of Nuclear Instrument Penetration 391 - Two compensated ion chambers were used for servo control of the regu- ,latlng rod and for some 1nterlocks. The power was recorded on a linear recorder. Seven decades were covered.by means of manual ‘range selector ,sw1tches. - One high sensitivity BF3 chamber was provided for power 1nd1cat10n during filling of the reactor. | 24.2.2 Process Radiation Instruments. p Two kinds of detectors, ion chambers and Geiger Mueller tubes, were used to indicate the level of: act1V1ty in various process streams and to provide necessary interlocks. 24,2.3 Health Physics Monitoring | - Gemma radiation wasvmonitored‘by'seven monitrons'located throughout the building. The air contamlnation was monitored for beta—gamma emitting -partlcles by seven constant a1r monitors. The building evacuation system operated when two or. more monitrons or two or more constant- air monitors from a speclflc group of. 1nstruments detected a high level of radiation or air contamination. 2k,2.4 Stack Activ:Lty Monitorlng The containment stack air was checked for beta gamma particulates by passing a side stream through a fllter paper andrmonrtorrngrthe activity .. with & GM tube. After'passing through the filter paper, the gaseous sample passed through a charcoal trap which was monitored by another GM tube to | detect 1od1ne. A second side stream passed through another filter paper. '_*ThlS was monltored for alpha partlculates using a thellium-actuated zinc sulfide screen detector. o ' "2h 2, 5 Temperature Detection o , - . One thousand seventy—one chromel-alumel thermocouples (some of which were installed spares) were prov1ded for temperature 1nd1catlon. A1l of ' the salt thermocouples except seven were located on the outside of the 1ines _p,or equlpment. The seven 1n—thermocouple wells vere located as follows: - one in the reactor neck one and two spares in the radlator inlet line; '-3§and “one and two spares in the radlator outlet line, - The more important 'ftemperatures were disPlayed on. S1ngle or multipoint recorders or indicators, ~ Readout of others was accomplished by meens of a scanner system'whlch al- lowed the signals from up to 100 thermocouples to be sent to a rotating 392 ‘mercury switch and subsequently to an oscilloscope for display. A switch allowed selection of any one of five groups of 100 signals. 24,2.6 Pressure Indicators | Most of thé remote indicating pressure and dp instruments were pneu- matic or electric force elements. When containment was required; the vents were referenced to atmospheric pressure through”roiling disphram seals. Strain gage and Bourdon pressure gages were used extensively. Small changes in reactor cell pressure were determined using a “hook gage". This was es- sentially a'water'manomEter w@th & micrometer for accurately reading changes in water level. ‘ | ' 24.,2,7 Level and Weight Indicators | Bubbler type level instrumentsrwere,used‘for measuring the salt levels in the fuel pump, coolant pump, and overflow tank. A fioat—type level in- strument was also installed in the coolent pump. The drain tanks were sus- pended by pneumatic weigh cells to determine the amount of salt ‘that they contained. In addition to this, two.resistant-type,single;point level probes were provided in each drain tank.. - Sight glasses, floats, bubblers, dp cells, etc., were used in the auxiliary systems. 24.2.8 Flow Méasfirément;f No flowrinstrument wes provided in the fuel salt loop, however & ven- turi meter was installed in the coolant salt loop. This was a standard venturi with a NaK-filled dp cell. Flows in auxiliary systems were meas- ‘fi:ed by‘orifices, capillaries pitot tubes, rotameters and matrix type flow | ;eiéments. | 2h;2.9 Miscellaneous A semi-continuous mass spectrometer was used to monitor the coolant ~air stack for beryllium. Other instrumentation included the following: pulse-type speed elements, ammeters, voltmeters, and wattmeters for the salt pump motors; potentiometer and synchro position indicators for the - control rods, fission chambers, and radiator doors; oxygen and moisture analyzers for the helium cover-gas system; and an oxygen analyzer for the cell atmosphere, 393 24,3 Initial’Checkout and Startup Tests A comprehensive functional checkout of the control, safety, and alarm instrument was made by Instrument and Controls personnel prior to operation fof each system, Although some design and wiring errors were found, these -were-of minor nature_and were easily corrected, In general the quality of installation was excellent.; The following were included in this checkout: the setp01nts of all switches were adjusted to the proper values; contin- uity and resistance checks of all thermocouples were made “the location of .each thermocouple was determined by heating the thermocouple and measuring the voltage at the patch panel' and the continuity of all circuits was , checked and all recorders vere put into operation. Most of this was done as construction was completed or as the instruments were put into service. Prior to the first circulation of flush and coolant salts, e complete ',operational check was made of all instruments. scheduled to be used, This was done follOW1ng the- instrument startup check list.22 This involved: (l) & complete checkout of all c1rcuits. This was done by changing the variable or inserting a false signal at the primary.element and assuring that all circuitry functioned at the proper setp01nts- (2) a complete check of the patch panel to assure that all thermocouples were connected to the proper readout instrument and that the operational records were up to date, (3) a complete test of all standby equipment to assure that it would start if needed and would function properly; (4) a complete inspection to assure 7_~¥that no switches were inactivated or Jumpers installed; and (5) tests to "assure that critical equipment fUnctioned properly, such as rod drop times. o - This instrument startup check list was repeated approximately every year.- Early- tests revealed errors in wiring, setpoints which had drifted and other instruments which did not function properly. Later tests 1ndi— pcated setpoints which needed to be reset and occasionally a malfunctioning ljinstrument. - Most of the troubles which were discovered.by doing the instrumentation ",pnstartup check lists were corrected immediately. Therefore_records are in- ’iadequate for statistical analysis."__.-“ 394 24l Periodic Testing Due to the importance of some instruments or ciréuits; they were also fiested periodicelly during operation. As indicated by the sections which 'fbllow,.these tests did not reveal many serious troubles which needed ¢or— rection. They did provide assurance that the circuits should function if heeded. In determining the amount and frequency of testing, careful con- sideration.should be given to the above'weighed against the harm done to equipment due to repetitive testing and the possibility of interruptibfi of operation° Rod scrams, load scrams, and reactor'drains occurred at the MSRE due primarily to testing of equipment or instrumentation. ' The periodic tests made at the MSRE and the results of these testé ~are given below. | | "2h.4,1 Nuclear Instruments A complete check of all nuclear instruments was made each month during nuclear operation. These were done by Instrument and Contfols personnél using Section 8A of the Operating Procedures.22 All tests were satisféc- tory except those listed in Table 2h.1. | - - 2k.4.2 Process Radiation Monitors The process radiation monitors were tested each week during operation by inserting = éource near thé primary detector and noting that each con- trol interlock functioned properly and that all'annunciations_occurred. All tests were satisfactory_except those listed in Table 24.2. Some early difficulty was encountered in that the hole in lead shielding'for inéerting the source was not properly placed. oh,4.3 Personnel Radistion and Stack Activity Monitoring Periodic tests of the health‘physics monitors and staék aétivity in- struments were originally done by MSRE personnel per Operating Procedure 8C.22 fThis was later taken over by other ORNL groups. The briginal séhe— dule for the health physics monitors was to make a source check‘of-each " instrument weekly, a matrix check monthly, and a complete evacuation test semiannually. As confidence in the instruments was established, the fre- quency of these tests was reduced to monthly, quarterly, and semiannually. 395 Table 24.1 Results,cf Periodic Tests of Nuclear Instruments Date Repairs_Necessary 3/22/66 Replacéddccndenser in the power suppiy of linear power - channel*Nc.*ls - 7/19/66 Replaced perlod balance module of nuclear safety channel NQ. 1. 16/16/677' Changed out high voltage supply of nuclear safety channel No. 3. T/1k/67 - Repaired scalerrof wide-fangeiCbunting channel No. 1. 8/12/68 Replaced reley of nuclear safety channel No. 3. 1/7/69 | Replaced fast trlp comparator of‘w1de-range counting ' channel No. 2. 2/3/69 Replaced chaflber of wide-range counting channel No. 1. 9/11/69 Replaced scaler of'W1de-range countlng channel No. 2. 10/21/69 Replaced operational amplifler of wide—range counting channel No, 1. 10/20/69 Replecedéeceler'of wide-range counting channel No. 1. 396 Tableazh.2 Results of Periodic Tests of Radiation Monitors Troubles Detected Date 4/13/66 Alerm did not occur, RM-596. 5/10/66 Instrument would not calibrate, RM-55T. 1/9/68 Two indicator lights burned out, RM—565,, 1/26/68 Indicator light burned out, RM-675. 7/21/68 Indicator light burned out, RM-82T. 3/3/69 Indicator light burned out, RM-565. Indicator light burned out, RM-565. 10/10/69 397 The stack monitors were originally tested with a source each week. Due to their excellent performance record, this was later reduced to monthly tests. - - ‘ fl Troubles encountered were repaired immediately. 2k, h.h Safety Circuits Periodic checks were made of all.circuits and instruments designated by the Instrument'ano'Cofitrols=group as being safety. These included rod scram-cirouits,,fuei pump and_overflow.tank'pressures.andzleVels, helium supply pressures, emergency fuel drain'cireuits, reactor cell pressures, coolant pump speeds and flows, radiator temperatures and,sampler-enricher interlocks. These tests involved simulating a failure apd checking as much of the circuitry as possible without interrupting operations. At.first these were performedrweeklyijthen ell except the rod scram checks were changed tora‘monthly basis.flAll-inatruments functioned Satisfaotorily ex- cept those indicated in;Tablefléh.B.‘ 24,5 Performance ‘of the Nuclesr Safety Instrumentation These instruments proved to be very relisble. There were periods when an sbnormal number of spurious trips occurred. Many'of‘these‘weré believed to have origInated in faulty, vibration-sensitive relays-in commercial elec- tronic switches which provided the high temperature trip- 51gnals.r Another possible source was the chatterlngrof the relays which- change the sensi- tivity of the flux amplifierelin;thetsafety,cirCuits;'"Correction_of_the e chattering and- eliminatiofi'6f*noiee'producing components“elsewhere'in the.™ system reduced -the frequency of‘these to & very tolerable: level.: * The fast trip comparators were found to be inoperative if a sufflci- ently large signsal was applled to the input. A diode was added to the fast trip comparator modules to elimlnate this difficulty. Operation of-the relay matrix in the nuclear system generated con- siderable noise, maklng it dlfflcult to reset the safety—system channels. The resistor-dlode combination for damping the voltage 1nduoed by relay operation’ wes replaced by & Zener diode and a diode comblnatlon ‘that was more satisfactory. 398 Teble 2k.3 Results of Periodic Tests of Safety Circuits Date Troubles Detected 5/2L /65 Noted that PR-522 and PI-522 did not agree. - 6/9/65 _Two safety channels tripped when testing one chennel, scram setpoint was at 120% instead of 150% and low current test _ did not function properly. | 1/15/66 Fuel pump pressure switch setp01nt needed resetting. h/10/66 Reactor cell pressure switch setpoint needed resetting. 5/20/66 -Two sampler-enrlcher pressure switches setp01nts needed resetting. 6/12/66 Sampler-enricher pressure switch needed resetting¢ 6/15/66 Rod motion was Jerky. 9/19/66 Two sampler;enricher-access door latches did not function properly. 11/2/66 A safety chanuel would not reset. 12/16/66 Reactor cell firessure switch setpoint needed resetting. 11/27/67 ‘Burned-out ‘indicator 1ight in safety,ehannel. | .5/13/69 Defective solenoid coil on reactor cell block valve header. 9/8/69 . Reactor outlet safety interlock would notclear. 11/24/69 - Safipler-euricher pressure switch setpoint needed resettiné. 399 In the summer of 1967, the 1-kW L8-V-dc to 120-V-ac inverter, which supplies ac power to one of the three safety channels, failed during sw1tch1ng of the 48-V de supply. It was repaired by replacement of two power transistors. During this same period the output of the ion chamber in safety channel 2 decreased drastically during a nonoperating period, and the chamber was replaced prior to resumption of operations. The trouble proved to be a. failure in a glass. seal that allowed water to enter the mag- nesia insulation in the cable, which is an 1ntegral part of the chamber. A period safety amplifier failed when llghtning struck the power line to the reactor site, and a replacement amplifier failed as it was being in- stalled, The field-effect transistor in this type oféamplifier was sus- ceptible to damage by transient voltages, and it was found.that under some ~conditions, damaging transients could be produced when the amplifier was removed - from or inserted 1nto the system. A protective c1rcuit was de- signed, tested and 1nstalled The module replacement procedure was modi- fied to reduce the poss1bi11ty of damage 1ncurred on 1nstallation of the module., Two relays in the safety relay matrices failed both with open. coil oircuits. A chattering contact on the fuelupump motor current relay caused safety channel 2 to trip several times before the problem was over- come by paralleling two contacts on the same relay. A defective SW1tch on the core outlet temperature also caused several channel trips and one re- actor scram before the trouble was identified and the SW1tch was replaced. A w1r1ng ‘error ‘in a safety circuit was discovered and corrected | Interlocks had recently ‘been added in the'"load scram. channels to drop the load when the control rods scram._ A wiring design error resulted in these interlocks - ”being bypassed by a safety Jumper. Although the c1rcuits were W1red this '-way for a time before. being discovered the scram 1nterlocks were always operative ‘during power operation, 51nce the reactor cannot go into the '"Operate, mode when any . safety Jumper is inserted. ' " Late in 1965, the power’ supply to the model q-2623 relay safety ele- ~ ments was changed from 115-V—ac to. 32—V—dc. ThlS was done to eliminate .ac pickup on the other modules through which the relay COll current was routed, The 115-V—ac relays were not changed at this time. These func- tioned satisfactorily until the-summer of 1967 when two of the 15 relays failed. By the'end of l967;mseven,had failed, all with open circuits or_ 400 safe conditions. In April 1968, all 15 were replaced with relays designed for 32-V-dc operation. Soon after installation, 3 of these failed due to contact welding (unsafe condition). This was apparently due to early | failure of defective relays. There were no more failures during.extensive tests made at this time or during subsequent operation. Starting Septem- ber 30,.1968, daily tests were conducted on the entire rod scrams relay matrix to detect single failures. (The matrix had been tested weekly be- fore that.) Noise,suppressors were installed across some of;the non- safety contacts of these relays to alleviate the noise which had sometimes caused difficulty during in-service tests. 24,6 Performance of the Wide-Range Counting Chennels Initial'criticality tests disclosed that the neutron flux attenuation in the instru:hent penetration _.did' not follow an -i’dea..l exponential curire. The deviation was too large to be sdequately comnensated for by the verni- stats in the wide;range counting'channels. This is illustrated byrcurve ‘A of Fig. 2L.2 which shows fission counter response (normalized count rate) vs withdrawal from the lower end of the penetration in guide tube 6. It was reasoneble to conclude from.this, end from similar curves, that the excess neutrons respongible for the distortedrpart of the curve were en- tering the penetration aslong its length. The count rates in. other guide tubes nearer the upper half of the penetration were even more distorted than curve A. Since a flux field with attenuation per Curve A precluded successful operation of the wide-range counting 1nstrumentation, shields of sheet cadmium were inserted in guide tubes 6 and 9 to shield the flSSlon chanbers from stray- neutrons. Curve B of Fig. 2h 2, normalized count rate ' vs distance, shows the improvement for guide tube 6. o | ~ Reactor period: infbrmetion from the wide—range counting channels 1n— hibited rod withdrawal and caused rod reverses.} Reactor period signals from counting channels operating at low input levels are characterized.by " slow response. This inherent delay produced a problem'w1th servo—controlled rod withdrawal during start. In & servo-controlled start, the demand sig- 'nal‘caused the regulating rod to withdraw until the period—controlled."W1th— drew inhibit" interlockzoperated. If the period continued toidecresse, the c NORMALIZED COUNT RATE -10° MSRE GUIDE TUBE NO.9 - - COUNT RATES IN GUIDE TUBE ' ' { A -~ NO.6 BEFORE AND AFTER ‘ _ 0! ' 'ADDING CADMIUM SHIELDING ‘ CONCRETE FILL g &’ ' : ' = ‘ N\ & SECTION THROUGH i~ REACTOR SHIELD— | < NUCLEAR INSTRUMENT _ - \ ' ' . PENETRATION w—a | \ \\ l ' ) ’.I"{ . F— ' ' 57 - - \ . \ \ GUIDE TUBE w3kl X - - \ ’ A * NO.6 r ELIN] A \ | ‘ & - \ SN - &0 REACTOR - SR e - N N : _ \ : : g CONTAINMENT ol N Ll 'MIDPLANE | 5in ‘ 4 CELL: =T ‘SR OFCORE | § ‘ , = \ ' ‘ = — ) FISSION COUNTER 1073 |=— CURVE “B" COUNT \ \ _ \ , l ]““m i A . © = RATE WITH CADMIUM \ ‘ o "’ i 7 |2 SHIELD ‘ ] } w . S - ¢ N . ' . REACTOR oL - — CURVE "A" COUNT RATE— VESSEL , = " WITHOUT CADMIUM = SHIELD ) - CADMIUM SHIELDING 1077 L o : SUBASSEMBLY -~ TYPICAL ) 20 40 60 0. IN' GUIDE TUBES THIS LENGTH 100% x, DISTANCE WITHDRAWN (in.) 6 AND 9 _ ( CADMIUM WRAP - e _ 10f-0in. ' 1011-Qin. ‘ SECTION A-A r__}'i o | : LOCATION OF _ - : : GUIOE TUBES , = =3 L3 O v{l (:)‘ | OUTER SHELL REMOVED FOR CLARITY—SHADED WEDGES ARE CADMIUM —UNSHADED WEDGES ARE ALUMINUM Fig. 24.2 Guide Tube Shield in the MSRE Instrument Penetration ORNL-OWG 66-2741 . o~ ) ot 402 "reverse" interlock acted to insert the rodslin_direot opposition to the servo demand. These "withdraw inhibit" and "reverse"'trip'points were originally established at periods of +20 and +10 sec resPectively. The ‘delayed low-level response of the "inhibit" ioterlock allofied sufficient incremental rod withdrawal to Produce a lb;sec period-and thus cause a re- verse. The ‘situation was aggravated by coastlng of the shim-locating motor in the regulating rod limit switch assembly. To correct this, ‘the "withdraw inhibit" and "reverse" period trlp poxnts were changed to +25 and +5 sec, respectively, an electro-mechanical clutch-brake was inserted in the shim- locatlng motor-drive train and dynamic breking circuitry was installed for the regulating rod drive motor, | Throughout operetion, dlfflculty was experienced with m01sture pene- tration into the fission chambers. The cause wes diagnosed as excessive strain and flexing of the tygon)tubing sheath on the electrical cables. 'I'he avera.ge lifetime was about 6 months prior to July 1968. At this time the type of tygon tubing used to cover the ceble and the method of sealing the chamber connections were .changed. This seemed to improve the moisture re31stance. : : ~ In early 1967, a failure occurred due to a short in the cable to the : preamplifler. The drive tube unit was modified to provide for controlled ceble bends and there were no reoccurrences .of this type of difficulty. 24,7 Performance of the_Linear Power Channels The linear power channels and rod servo instrumentetion perfOrmed very well throughout the operation. When the fuel was changed to'233U, the rod- control servo was modified to allow an increase in the servo dead band to . compensate for the more rapid flux response of the reactor. The compensated ion chembers use & small electric motor to change com- pensation. Early in 1966 one of these motors had to be replaced. A water leak occurred in one of the compenssated ion chambers in the summer of 1969. When it was examined, humerous leaks were found along the seam weld in the 321 stainless steel bellows sheathing the cable, and the aluminum can at the outer support'ring was"honeycoflbedioy eorrosion. (The water in the instrument shaft contained lithium nitrite buffered W1th boric acid for inhibition of corrosion.) 403 2h.8 BFz Nuclear Instrumentation Because of very-unfavbrablergeometry,'the“strongest practicablepneu- tron~source would not~prodmce72'counts/secu£rom the fission'counters in the wide-range counting chahnels until the core vessel was:approximately half full of fuel salt; neitherrwould itfproduce 2 counts/sec with flush salt in the core at any level- .This'was the minimum count rate required to obtain the permissive confldence 1nterlock which allows filling the core vessel and w1thdraw1ng the rods. Therefore a counting channel using a sensitive BF3 counter was added to estsblish "confidence" when the core ~ vessel was less than halfefilled with fuel salt. This was installed early - in 1966. The chamber had to be replaced in the summer of 1967 due to mois- ture leakage'into the cable. .No other troubles were encountered. 24,9 Nuclear Instrument Penetration ‘The:nuclear power produced at the‘MSRE was_determihed by an overall system'heat balance. All;nuclear power inStruments were calibrated to agree with this primary standard. Durlng extended runs at higher powers, the nuclear instruments 1nd1cated 15 to 20% higher than the heat balance. Th1s was found to be due to a rise in ‘water temperature in the nuclear instru— ment penetratlon,which apparently changed the attenuatlon characteristics of the water. In June 1966, & heat exchanger system was placed in operation - to cool the water which reduced the temperature change from zero to full | power to A18°F -from 72°F and reduced the difference in the two power meas- _'urements to about 57 - 24,10 Pé&férman¢é°6f*£fié Process Radiation Monitors . The process radiatlon monltors proved very rellable.p,No-radiation ._elements had to be’ replaced durlng the entire operation. Occasional re- ;:pairs were necessary on .the electronics. _One rod end load scram resulted from a false signal from- one of these monltors (RE-528). 404 24,11 Performance of the Personnel Radiation = - ._Mbnitoring,andrBuilding Evacuation System During early testing it wes found that there were some areas where the building evacuatlon horns could not be heard Two addltional horns and four addit10nal beacon alarm 1ights were installed Other than oc- ca31onal minor repairs, the system has functioned satisfactorlly. 2h.A12 . Performance of the: Stack Monitoring System The . stack mbnitoring sysfem.was very relisble.- The manuallrange _-SW1tching made it difficult to 1nterpret date from the recorder charts ‘and the: Rustrak recorders were very inconvenient to use.. 2h 13 Performance of Thermocouples-end the- . Temperature Readout ~gnd~Control- Instrumentation A total of 1071 thermocouples were installed at the MSRE. Of these, 866 were on salt systems (351 on the. circulating loops and 515 on the drain tanks,ydrain-lines, end freeze vaelves). Of the 1071, only 12 have failed in five_years.of-serviee; 'Five others were damaged during construction and maintenance. A breskdown of the failures is given in.Table 2L.l, Table 24.U4 Failures Among the 1071 MSRELTflermOCOuples .. Nature of Failure 2 ¢ .. Number Damaged during construction Damaged,duriag maintenance' Lead ofiefi-during operation Abnormally low reading (detached?) Unknown reason Ww w o N w T ' Total | AT 405 Only three thermocouple wells were provided in the c1rculat1ng salt systems " one each in the coolant radiator inlet and outlet pipes ‘and one in the reactor neck. The remalning thermocouples vere attached to the pipe or vessel walls. The thermocouples on the radiator tubes were insulated ‘to protect them from the effects of the high-velocity air that flows over '”them during power operation; the others were not insulated and thus were subject to error because of exposure to heater shine and to thermal con- vection flow of the cell atmosphere within the heater insulatlon. In March 1965, with the fuel and coolant systems circulating salt at isothermal condltlons, a complete set of readings was taken from all the thermocouples that should reed the temperature- of the circulating salt, A similar set of data was taken in June 1967 at the start of Run 12, The results of the two sets of.measuremehts are shown in Table 24.5. Comparison of the standard ~deviations for the radiatorrtherm000uples‘with those 'for the other thermo- couples shows the effect of_inSulation on reducing the scatter,,;Comparison of the sets of data taken over two years apart shows very little\change, certainly no greater scatter. -Figure 2h.3 shows that the statistical dis- tributioniof the deviations on individual thermocouples from the mean also changed little in the two years. - Table 24k.5 Comparison of Readings of Thermocouples of Salt Piping and Vessels Taken with the Salt Isothermal h™ 8 Indlcated Temperature (°F) Thermocouple - .. Location '.fl'March 1965 © June 1967 "rRadiator tubes - 'll62;6 % 6.7f o _1208.5 # 3 3 Other - - 1102.1 % 13.0 - 1206.7:+-12.3 AL - 1102,3 % 10,6 . 120T.4 £ 9.8 406 ORNL-DWG 67-41786 100 ¢ , — fg' o Te—® o - o5 R ‘ O O 90 o g & < = 80 5 g o o . . a 0 ° Lt I U ® DATA TAKEN - g 60 ~ MARCH 1965 5 - o DATA TAKEN ° a | JUNE 1967 w 50 ' - ' o W S | [ o 40 < _ L m § S 30 e i | o E 20 —8 o : 0 & . .. . ’o 10 3 . . Q " - o485 Ole—L o ov% -60 -40 -20 O 20 40 €0 DEVIATION FROM AVERAGE (°F) Fig. 24.3 Comparison of MSRE Thermocouple Data from March 1965 and June 1967 407 The scatter in the various thermocouple readings was reduced to an acceptable level by using biases to correct each reading to %he overall average measured while both fuel and coolant systems were circulating salt at isothermal conditions;“rThese biases were entered into the computer and were automatically applied to the thermocouple readings. The biases were revised at the beginning of essh run and were checked when isothermal con- ditions existed during the runs. Generally the biased thermocouple readings were reliable, However in a few cases, there were shifts which caused cal- culational errors. - During early operation the thermocouple scanner gave considerable dif- ficulty due to 60~cycle noise pickup, poor stability, and drifting of the salvaged oscilloscopes. Refinements were needed in design to allow better identification of scanner'poifits and provide a means by which the operator could calibrate the instruments. After these were corrected, the system operated very satlsfactorlly. The rotating mercury switches lasted much 7'longer than'the expected 1000-hour mean llfe. One switch failure occurred when the nitrogen purge gas was 1nadvertently stopped. | | Single point Electra Systems alarm switch modules were used for con- trol of freeze valves and for other alarm and control_actlons, These gave considerable trouble during early operation due to drifting or dual set- points and general maloperafidn. A number of mo&ifications were made to correct these. Printed circuit—board contacts were gold-plated to reduce contact resistanee, the trim pots used for hysteresis adjustment were re- placed with fixed resisfiofs; and resistor values in modules hafiing ambig- - uous (dual setp01nts were . changed to restore the proper blas levels. These changes, together w1th stabllizatlon by aging, of crltlcal resistors in ,the SW1tch modules and more :1gorous perlodlc testlng procedures<1mproved the performance. AAcheck sh6fie§ that out of 109 switch setpoints, 83% had shifted less than 20°F Q#er sfsix-month'period. Multiple setpoints still occurred and various other*faiIUres were encountered. During'1968 and 1969, records were kept on . the fallures of 1nd1cator llghts on these mo- dules, There were about 50 fallures per year. ‘This was 1mportant because & burned-out llght bulb could cause alarmror control actlon. . S e N 408 24 .14 - Performance of PressurewDetectors_{ Most . of the pressure instruments at the MSRE performed very well, Dif- ficulty was- encountered with the differential pressure cell used to obtain the pressure drop in the helium flow through the charcoal beds (PAT-556). The span end zero settings shifted badly although the pressure capsbility had never been exceeded. Three dp cells failed in this service. Two of Aé}these were removed, tested, and inspected without determining the cause of " the trouble._ The last’ replacement functioned satisfactorily at first but then gave similar difficulties. It is still instelled: _2&,15' Performancenof.Level Indicators The: bubbler-type level instruments used in the fuel pump overflow tank and coolent pumps performed well. More details are given in. Sections 5 8 and 6.5. Some difficulty was encountered in controlling the purge flow un- \til the throttling velves vere replaced,- A high and. 16w level resistance type level probe was provided on. each (i} drain tank. During early operation the excitation and signal cable leads on both probes of the fuel flush tank failed. These failures were caused by - excessive temperature which caused oxidation and embrittlement of the copper-clad mineral—insulated-copper—w1re cables. These cables were de— signed on the assumption that they would.be routed in- eir. above the tank insulation and that their operating temperature would-not exceed 200°F- however, in the actual installation, the cables were covered with 1nsula—' tion and the temperature at the point of attachment to the probe was prob- ably 1n excess of 800°F Repairs were accomplished.by replacing the copper- - ‘cled, mineral-insulated copper wire excitation and - signel cables and por- “tions of the probe head assembly with a stainless-sheathed, ceramic-beaded nickel-uire cable assembly. In the summer of-1968, one of the probes in fuel drain tank No. 1 failed. The failure was found to be an open lead w1re inside the cell The probe-wasrrestored to service by a cross con- nectionvouts1de the cell to the equivalent lead of the other probe,' | The’ initial 1nstrumentation provided to assure proper Water level in the ‘vapor-condensing tank consisted of Ik resistance type prdbes spaced ‘ U et bt s brdnnon gl @ mre agigg ere 409 h in. apart near the desired water level. In September 1966, while doing ------ the instrumentation startup check 1list, one of the two high-level sw1tches was found to be defective.j Loss_of this switch caused the loss of one‘chan-" ‘nel of information needed to'establish'that the water 1evel in the tank was correct ' Since a second switch failure might require that the reactor be shut down until the sw1tches could be repaired, and since the removal of Vthe SW1tches from the tank is a difficult operation, a bubbler-type level : measuring system, which 1ncluded a containment block valve and associated safety circuits in the purge supply, was designed and was installed in the ~reactor~-cell vapor SUppre381on tank. This installation utilized 8 dip tube vhich was included in the original design in. anticipation'of such need. This new level system also enabled the operator to check the water level in the tanks as a routine procedure. ol 16 - Performance of the Drain Tank Weighing Systems The same type pneumaticiueighing devices were used on the fuel.drain .tanks and the coolant drain tank' ‘The coolant drain- tankrweight indicators ' proved to be stable, show1ng no long—term drlfts or effects of external 'variables.- With 5756 1b of salt in the tank at about l200°F the extreme spread of hO 1ndicated weights over a period of a week was + 22 1b. This was only 0. h% and was quite satisfactory. o ' The 1nd1cated weights of ‘the three tanks in the fuel system exhibited rather large unexplalned changes.- In some cases these amounted to 200 to ,300 pounds. The mechanism causing this was not definitexy established but probably was due to changes in forces on the syspended tanks as tempera- _tures of attached piping and the tank furnaces changed. Reactor.cell pres- sure seemed to also affect the readings. The weighing systems were useful in observing transfers of salt and for filling and draining the reactor. In addition to this calibration drift difficulty was experienced with " the multiposition pneumatic selector switches. Manometer readout was ac- ”:complished by selecting a particular weigh cell channel‘uith pneumatic'se- tr_lector velves. The valves were composed of & stacked array. of indiV1dual valves operated by cams on the operating handle shaft. Leaks 'in these valves gave false weight indications. A redesign of the switching device solved this problem. 410 Prior to power operation, one of the weigh cells failed and was re- placed. " The failure was determined- to be due to p1tt1ng of the baffle and " nozzle in the cell, This pitting was apparently caused by amalgamation of | mercury with the plating on the baffle and nozzle, How the mercury got in- ,‘to the weigh cells has not been determined however, it was believed to have come from the manometers and to have been precipitated on the baffle | by expansion cooling of- the air leav1ng the n_ozzle° Apprec1able quantities of merciry were also found in the.tare;preSSUre regulators on the control panel; however, no mercurypwas'found'in thedinterconnecting tubing_or in other portions of the system, 24,17 Performance of the Coolant Salt Flowmeters | Several days after the start of coolant salt circulstion, the output of one of the two salt- flowmeter channels started drifting down scale. The output of the other chpnnel remained steady. The trouble was-isolated to a zero shift and possibly a span‘ahift in the NaK-filled differential pres- sure tranSmitter in the drifting channel Since the exact cause could not be determined e spare dp cell was installed. Both channels functioned satisfactorily throughout the remainder of the reactor operations,; Tests on the defective unit were inconclusive- however,_ it was determinedithat the shifts were temperaturewinduced zZero shifts possibly caused by incom- plete f£illing or a leek in the - silicone oil portion of the‘instrum.entq An operational inconvenience per51sted “throughout’ operations° When- ever the radiastor air flow was increased, air leaking through the 1nsula~ tion around the salt legs to the dp cells caused the temperature to decrease 1rap1dly. This necessitated adJustment of the heaters. | Due to the discrepancy between the reactor power level 1nd1cated by fuel burnup, ‘heat balance ete., it is planned to recalibrate the dp cells dnring the next fiscal year. 411 2,18 Performance of Relays The difficulties encountered with the control red relays:are described in Section 24.5. Experience with other relays is given below. After ebout 2 years of operation, the 48-V-dc-operated relays showed ~considerable heat damage to their bakelite frames. The manufacturer, General Electric, advisedvthat.overheating of this particular model was a common problem if the relays were continuously energized. Early in 1967, twenty of the 139 relays were replaced with a later improved model. Within a few months some of these also showed signs of deterioration. Therefore in June 1967, all 139'relayS‘wefe field—mbdified'byrreplacihg the built=in resistors with externally_mounted_resistors.-‘No trouble was experienced after this modlflcation. In September 1969, the load-scram c1reu1t » Which drops the radiator doors and stops the blower, tripped several times. Investlgatlon showed that some of the relasy contacts had developed unusually high resistance due to oxide films, Because of the way the contacts were paralleled in the matrix, the film was not Burned off each time the contact closed, as in a fiormei application. These,contaetsrwere cleaned and no further dif- ficulty was encountered. 24,19 Training Simulation Two "on~site" reactor kinetics simulators were developed for the pur- pose of training the MSRE operstors in nuclear startup and power operation. The startup simulator -used the control rod position signals as inputs, and provided outputs of log COunt'fateg period, log power, and linear power, The resctor's period interlocks, flux control system, and linear flux range selector were also operationel ~ In addition-to this, the power level simu- :lator used the radiator door p051t10n and cooling air pressure drop signals as .inputs and provided readout of key system temperatures.. Both simulators were set up on general purppse,_pprt&ble EAT TR-10 analog computers. Much of the actual MSRE har&ware;feuch-as control rods were used rather than simulated. Thus the operatofs manipulated the actual reactor controls and became used to the instrumentation and controls system. This proved to be a very relisble training tool. 412 2k.20 Miscellaneous The original mass spectrometer used to monitor for beryllium in the coolant stack was replaced with an improved instrument near the start of power operation. Onlyroccasional repeirs or preventative maintenance was required since then. | | | : | | Fdlded-charts were used on some recorders.- Thése did not function properly due to the low chart speeds being used. 24,21 Conclusions andaRecommendations _.'Considering the quantity and complexity of the instrumentation, a mihimal amount -of difficulty was encountered. Since the MSRE was an ex- perimen€al réactor; it was in many areas over-instrumented. This was proba- bly due to not -knowing what information might be needed and not taking enough credit for the gbility. of the operators. This led to unnecessary difficulties in normal operstions or in running special experiments. At the same time, there were areas where additionsl informetion would have been beneficizl. - . 1'0 3. 4. 5, 6. 7. 9. 10. 11, 12. 13, 14. 15. 16. s - and Foam in the MSRE 0RNL—TM+3027 (June 1970) : 719' 20, 21, 413 REFERENCES M. W. Rosenthal, P, R, Kasten, and R, B, Briggs, "Molten-Salt Reac- ~ tors —-History, Status, and Potential, " Nucl. Appl Tbch., 8, 107 (1970). | MSR Program Semiannu. Progr. Rep., Jan. 31, 1964, ORNL-3626. MSR Program Semianmu. Progr. Rep., July 31, 1967, ORNL-3708. ' MSR Program Semiannu. Progr. Rep., Feb. 28, 1965, ORNL-3812. MSR Program Semiannu. Progr. Rep., Aug. 31, 1965, ORNL-3872. MSR Program Semiannu. Progr.’ Rep., Feb. 28, 1966, ORNL-3936. MSR Program Semiannu.Progr. Rep., Aug. 31, 1966, ORNL-4037. . MSR Program Semiannu. Progr. Rep., Feb. 28, 1967, ORNL-4119. - MSR Program Semiannu. Progr. Rep., Aug. 31, 1967, ORNL-4191. " MSR ProgramVSCmiannufcRrogr,rR@p., Feb. 29, 1968, ORNL-4254, MSR Program Semtannu Progr.,R@p.,rAug. 31, 1968, RNL—4344. MSR Program Semzannu;Progf; Rep., Feb. 28, 1969, ORNL-4396." MER Program Semtannu.'Pfégr.flep.,_Aug. 31, 1969, ORNL-4449. MSR Program Semzannu.-Prbgf. Rep., Feb. 28, 1970 0RNL-4548. ‘R, C. Robertson, MSRE Design and Operations Report Part .-I — De- scription of Reactor Design, 0RNL-TMF728 (Jan. 1965) R. H. Guymon, P N Haubenreich and J.,R Engel MSRE Design and __Operations Report Part XI-— Test Program, RNL—TME 11 (Nov. 1966) TR. B, Lindauer, Processing of the MSRE Flush and Fuel Salts, ORNL— - TM-2578° (Aug.,1969).:~,;;g,, | Jo R. Engel P.-N..Haubenreich and A. Houtzeel Spray, Mist Bubbles P. N. Haubenreich and M Richardson, Plans for Post—Operation Exami- ' nation of the MSRE ORNL-TMFZ974 (April 1970). }C. H, Gabbard Reactor Power Measurement and Heat-Transfer Performance in the MSRE, 0RNL—TM~3002 (May 1970) - J. R. Engel MSRE Design and ‘Operations Report Part XI—A-— Test Program for 2*2U Operation, ORNL-TM-2304 (Sept. 1968) 22, 23. 24. 25. 26 * 270 28. 29. 30. 31. 32. 33. 34, 35, 36. 37. 414 - References (continued) R H. Guymon, MSRE Design and Operations Report Part VIII, Operating Procedures, Vols, I and II, ORNL-TM-908 (Dec. 1965). R. E. Thoma, Chemical Aspects of MSRE Operntions, ORNL-4658 (Dec. 1971). R. H. Guymon and P. N. Haubenreich, MSRE Design and Operations Report, Part VI, Operating Safety Limits for the Molten-Salt. Reactor Experiment, 0RNL—TM—733 (3rd revision), (July 25, 1969). R. H. Guymon, R. C. Steffy, Jr., and C. H. Gabbard, Preliminary‘Evalu- ation of the Leak in the MSRE Primary System Which Occurred during the Final Shutdown, ORNL-CF-70-4-24, (April 22, 1970). P. G. Smith, Development of Fuel- and Coolant~Salt Centrifugal Pumps for the MSRE, ORNL-TM-2987 (Oct. 1970). C. H. Gabbard, Inspection of MSRE Fuel Circulating Pump after the ‘Zero-Power Experiments (Run 3), ORNL-CF-66-8-5 (Aug. 1966). R. B. Briggs, Measurement of Strains in Heat Exchanger Nozzle and Piping of MSRE, ORNL-CF-66-2-66, (Feb. 1966). C. H. Gabbard, Thermal-Stress and Strain-Fatigue Analyses of the MSRE Fuel and Coolant Pump Tanks, ORNL-TM~78 (Oct. 1962). C. H. Gabbard, R. J. Kédl, and H. B. Piper, Heat Transfer Performance of the MSRE Heat Exchanger and Radiator, ORNL~CF-67-3-38 (Mar. 1967). C. H. Gabbard, Reactor Power Measurement and Heat Transfer Performance in the Molten Salt. Reactor Experiment, ORNL-TM-3002 (May 1970). H. E. McCoy, An Evaluation of the Molten Salt Reactor Experiment Hastelloy N Surveillance Specimens — First Group ORNL-TM 1997 (Nov. 1967). H. E. McCoy, An Evaluation of the Molten Salt Reactor Experiment Hastel- loy N Surveillance Specimens — Second Group ORNL-TM-2359 (Feb. 1969). H. E. McCoy, An Evaluation of the Molten Salt Reactor Experiment Hastel- loy N Surveillance Specimens — Third Group, ORNL-TM-2647 (Jan.‘1970). H. E. McCoy, An Evaluation of the Molten Salt Reactor Experiment Hastel- loy N Surveillance Specimens — Fourth Group, ORNL-TM-3036 (Mar. 1971). H. E. McCoy and B. MbNabB, Intergranular Cracking of iNOR—SCin the MSRE, ORNL-4829, (Nov. 1972). _ . C. H. Gabbard, Design and Construction of Core Irradiation—Specimen " Array for MSRE Runs 19 and 20, ORNL-TM-2743 (Dec. 1969). C O 415 'References (continued) -38. R. B. Briggs, Effects of Irradiation of Service Life of MSRE, ORNL- CF-66-5-16 (May 1966) : . , 39. R. B. Briggs, Assessment of Service Life of MSRE ORNL—CF—69 8-3, (Aug. 1969). | 40, B. E. Prince et al., Zero-Power Physics Experiments on the Molten-Salt Experiment, 0RNL-4233 (Feb 1968) ' 41. J, R. Engel and B. E. Prince, Zero~Power Experiments with 2°°U in the MSRE, ORNL-TM-3963 (Dec. 1972) _ 42. C. H.-Gabbard and C.;K,xMcGlothlan,-A Review of Experience with the MSRE Main Blowers, 0RNL—CF—67—4—1, (Ahg. 1967). 43, C. H. Gabbard, Bearing Failures of Main Blowers MB-1 and MB-2, ORNL~- : CF—67—4—1 Addendumrl (Mar. 1968). 44, R, L. Baxter and D. L. Bernhard "Vibration Tolerances for Industry," paper presented at ASME Plant Engineering and Maintenance Conference, Detroit, Michigan, Paper No. 67-PEM-14 (Apr. 10-12, 1967). 45. E. C. Parrish and R, W. Schneider, Tests of High Efficiency Filters : -and Filter Installation at ORNL,; ORNL-3442, (May 1963). 46. H. R. Payne, The”Mechanical.Design of the MSRE Control Rods and Re- actor Access Nozzle, MSR-61-158 (June 1961). 47. J. L. Crowley, Temperature and Salt Levels in the MSRE Reactor Access Nozzle — A Possible Relation to MSRE Power Blips, MSR-69-13 (Feb. 1969). 48, P, N, Haubenreich, R. Blumberg, and M. Richardson, "Maintenance of the - Molten-Salt Reactor Experiment'" paper presented at Winter Meeting, ANS, Washington D. C. (Nov. 1970) . _ 49, C. H. Gabbard, Radiation Dose Rates to the MSRE Fuel—Pump Motor and Control Rod Drives, MSRF65“48 50. R. C. Robertson, MSRE Design and Operations Report, 0RNL-TM—728 51. R, H. Guymon, MSRE Design and Operations Report, Part VII Operating Procedures, ORNL"TMFQOB Vol. II (Dec. 1965) . 52, R. B. Gallaher, MSRE Sampler—Enricher System Proposal ORNL-CF-61-5- 120, - (May 1961). 53, MSRE Staff, MSRE Sampler-—Enricher, An Account of Recent Difficulties ' and Remedial Action, MSR~67«73 (Sept. 1967). 54- 55.". 56. 57. 58. 59. 60. 61. 62. 63. 416 References (cpntinued) P. N. Haubenreich and MSRE Staff, An Account of Difficulties with the Sampler-Enricher that Led to a Second Capsule Being Left in the Pump Bowl, MSR-68-125 (Sept. 1968). S. E. Beall et al., MSRE Design and Operations Report Part V, Reactor ‘Safety Analysis Report, ORNL-TM-732 (Aug. 1964). R. H. Guymon, Tests of the MSRE Reactor and Drain Tank Cells, MSRr 62—95 (Nov. 1962). | P. N. Haubenreich et al., MSRE Design and Operations Report Part III, Nuclear Analysis, ORNL-TM-730 (Feb 1964) . G. H. Burger, J. R. Engel, and C. Do Martin, Computer Manual for MSRE Operators, ORNL-CF=~67-1-28" (Jan. 1967) . P. N. Haubenreich, Safety Considerations in Resumption of MSRE Operation, 0RNL—CF—69—8—10 (Aug. 1969). 'P. N. Haubenreich, ReSponses to RORC Recommendations for MSRE, MSR- B. F. Hitch et al., Tests of Various Particle Filters for REmoval of &;fi 0il Mists and Hydrocarbon Vapor, ORNL-TM-1623 (Sept. 1966). W.VC.-Ulrich, Purge of MSRE_Off—gas Line 522, MSRr66—8; (Mar. 1966). R. H, Guymon and J. R. Engel, Introduction of Oxygen in Fuel-Drain Tank, MSRr67—1 (Feb. 1967). 84-85. 1. 2. 3. 5. .6. 7. 9. 10. 11. 12. 13. 14-18, 19. 20. 21. 22. 23. 24, 25. 26. 57+58. . 50-60. 61. 62, 63. 64. 54-56. 417 ORNL-TM~3039 Internal Distribution S. E. Beall o 27. R. N. Lyon E. S. Bettis o -~ 28. R, E. MacPherson R. Blumberg o - 29. H. E. McCoy R. B. Briggs : : 30. H. C. McCurdy. W. B. Cottrell - | 31. A, J. Miller J. L. Crowley S ' 32. R. L. Moore F. L. Culler ' 33. L. C. Oakes S. J. Ditto- : o 34. A. M. Perry J. E. Engel 35. M. Richardson D. E. Ferguson o - 36-37., M. W. Rosenthal A. P. Fraas ' 38. Dunlap Scott C. H. Gabbard o 39. M. R. Sheldon W. R. Grimes 7 40. M. J. Skinner R. H. Guymon oo 41, I. Spiewak P. H. Harley o 42. D. A. Sundberg P. N. Haubenreich _ 43. J. R, Tallackson H. W. Hoffman o 44, R. E. Thoma T. L. Hudson , 45. D. B. Trauger P. R. Kasten : 46. G. D. Whitman A. I. Krakoviak 47-48. Central Research Library Kermit Laughon, AEC—OSR' 49, Y~12 Document Reference Section M. I Lundin : 50-52. Laboratory Records Department 53. Laboratory Records (RC) . External Distribution Director,.Divisionrof Reactor Licensing, USAEC, Washington D.C. 20545 " Director, Division of Reactor Standards, USAEC, Washington, D.C. 20545 N. Haberman, USAEC, Washington, D. C. 20545 D. F. Cope, AEC-ORO ‘M. Shaw, USAEC, Washington, D.C.. 20545 65. 66-82. - 83. David Elias,_USAEc,-Washington, D.C. 20545 A, Houtzeel, TNO, 176 Second Ave., Waltham, Mass. 02154 R. C. Steffy, TVA 303 -Power Building Chattanooga, Tenn. 37401 Manager, Technical Information Center, AEC (For ACRS MEmbers) Research and Technical: Support Division, AEC, ORO : Technical Information Center, AEC .