COMPONENT DESIGN FOR LMFBR'S

R. H. Fillnow,Project Manager, FFTF Project

Westinghouse Advanced Reactors Division

L. L. FranceManager, Materials Applications TechnologyWestinghouse Advanced Reactors Division

M. C. Zeiinvary,FFTF Pump Program Manager

Westinghouse Electro-Mechanical Division

R.OFox .Principal Engineer, FFTF Project,

Westinghouse Advanced Reactors Division

For Presentation at theAnnual Meeting of the American Power Conference

and Publication in the Proceedings of theAmerican Power Conference

April 2123,1975 »

Westinghouse Advanced Reactors DivisionThis paper is based on work performed for the Energy Research andDevelopment Administration for the Fast Fiux Test Facility

TMI<£ ^nruMFNT UNLIMITED

COMPONENT DESIGN FOR LMFBR'S

R. K. Fillnow,Project Manager, FFTF Project

And

L. L. France,Maneger, Materials Applications Technology

And

R. 0 . FoxPrincipal Engineer, FFTF Project

Westinghouse Advanced Reactors DivisionMadison, Pennsylvania

M. C. Zerinvary,FFTF Pump Program Manager,

Westinghouse Electro-Mechanical DivisionCheswick, Pennsylvania

- NOTICE-This report was prepared 2s an account of worksponsored by the United Stares Government. Neitherthe United States nor the United States EnergyResearch and Development Administration, nor any oftheir employees, nor any of their contractors,subcontractors, or theii employees, makes anywarranty, express or implied, or assumes any legalliability or responsibility for the accuracy, completenessor usefulness of any information, apparatus, product orprocess disclosed, or represents that its use would notinfringe privately owned rights.

Component Design for LMFBR's Conference Paper R.H. Fi l inow APC

Chicago, 111. 4 /21 -23 /75

C;3TR!BUT10;>! OF THIS DOCUMENT UfiuiVi!!EO\

LIST OF FIGURES

Figure Title Page

1 Core Basket Upended During Fabrication 10

2 Instrument Tree 11

3 Reactor Vessel/Closure Head and Main Support 12Structure Interface

4 Primary Sodium Pump 13

5 IHX Tube Bundle During Tube-to-Tubesheet Welding 14

6 In-Vessel Handling Machine 15

COMPONENT DESIGN FOR LMFBR'S

INTRODUCTION

Pursuant to a systematic development ofenergy independence in the United States,the development of a commercial breederreactor is considered a major part of thenational program to provide this indepen-dence. The commercial breeder reactorconcept is the most expeditious approach toa reliable, safe, economic, and environ-mentally attractive long term source ofelectrical energy. A-vital link in thetotal proaram is the design and constructionof the Fast Flux Test Facility (FFTF), Inaddition to providing :a; testing facil ityfor breeder reactor structural and fuelmaterial, the FFTF program has the goalsof developing: analytical techniques forhigh temperature plant design; cultivatingan industrial base for design and fabrica-tion of liquid metal reactor and plantcomponents; and, training personnel indesign, fabrication and plant operations.As a result of this program, there willexist experienced design and manufacturingstaffs who are well aware of the specialconsiderations which must be factored intothe design and fabrication of liquid metalfast breeder reactor (LMFBR) components.

The FFTF is a high temperature liquid sodium-cooled reactor fueled by a mixture of piuto-dioxide and uranium dioxide. Although FFTFcannot breed additional fuel, it does providethe thermal, hydraulic and neutron environ-ment of a commercial breeder reactor for thetesting of material and components. Compo-nents and plant are of such a size that FFTFcan truly be considered as a first vital stepin the evolution of the commercial breederreactor. The FFTF power level of 400 MWt iscomparable to that of the French Phenixreactor and the British Prototype FastReactor. Operating conditions such astemperature and pressure are comparable

to those planned for both the first demon-stration plant and the prototype commercialbreeder reactor. A full technical descrip-tion of the FFTF has been presented else-where in open literature.lr2 It is intendedto review in this paper the progress madein developing the design and fabricationcapability of LMFBR components.

DEVELOPMENT OF HIGH TEMPERATURE DESIGN METHODS

Prior to FFTF, the state of knowledge fordesign of liquid metal cooled reactorsoperating at elevated temperatures was quitelimited. The operation of LMFBR's, in par-ticular at temperatures above TOOO°F, requiredsignificantextension of mechanical designmethods used in the design of lower tempera-ture plants. Of the two most notable U.S.breeder reactor designs, EBR-II operates ata coolant temperature of 883°F and Fermi-Ihad a 800°F* operating temperature. Incontrast, coolant temperatures of existinglight water reactors are below 650°F. Asa result, the ASME structural design criteriaembodied in Section III, "Rules for theConstruction of Nuclear Power Plant Compo-nents," has concentrated on a lower tempera-ture thermal environment employing in themain, linear elastic analyses. For FFTF,the design temperature is 1050°F, and theonly ASME rules for elevated temperatureClass I components available in earlystageso." design were contained in the four pagesof Code Case 1331-4. In addition, theresponse and failure characteristics ofaustenitic stainless steel above 800°F werenot well defined.

*Fermi-I was designed for a 900°F operatingtemperature.

Structural design criteria developed duringthe past six years as part of the FFTF pro-gram includes failure mode effect analyses,stress limits which depend on the durationof load, and recognition and reliance oninelastic analysis. Extensive determina-tions of high temperature material propertieswere obtained from a variety of sources tosupport the analytical efforts. These newconcepts have led the ASME to expand CodeCase 1331-4 to over 100 pages, and morerecently, to adopt Code Case 1592 as theelevated temperature design guide forSection III components. Today we haveprogressed to a level of capability wherewe can perform the necessary analyses toinsure the safe and reliable design ofcommercial LMFBR's.

DESIGN AND EVALUATION OF MAJOR PLANTCOMPONENTS

With few exceptions, the design and manufac-ture of critical components for the FFTFplant were performed by suppliers with littleor no previous experience with high tempera-ture components. Although design of thecomponents,, in most cases, did not pose unduemanufacturing difficulties, fabrication ofstainless steel components in shops whereprimary emphasis had previously been basedon carbon steel component fabrication didpresent problems in handling, cleaning, andmachining. One area of particular difficultywas maintaining an adequate cleanlinesslevel in a shop where such levels werepreviously unnecessary. The relatively thinwalls of large vessels presented unique pro-blems in handling and precise machining tomaintain concentricity. Difficultiesencountered in achieving satisfactory weldquality caused significant delays in com-ponent manufacture. Considerable time andeffort was expended in achieving the requiredwelder and inspector qualifications neededto meet the rigorous requirements. Ingeneral, after capabilities were developedto achieve the necessary quality levels,welding proceeded on a satisfactoryschedule.

In addition, FFTF design introduced the appli-cation of Reactor Development and Technology(ROT) Standards. The RDT Standards setexacting requirements for a wide range oftopics such as materials, welding, materialforming processes, non-destructive testprocedures and acceptance criteria, clean-liness and quality assurance. These stan-dards supplement existing nationally recog-nized standards to staisfy more stringentsodium reactor technology, safety and

reliability requirements. Lack of familiaritywith these standards contributed to manufac-turing difficulties.

During the manufacture of FFTF components,significant advances were made in the weldingof complex shapes without incurring unaccept-able weld distrotion. Three such exampleswere the fabrication of the instrument tree,the core basket, and the core support struc-ture. In the case of the core basket (Fig-ure 1), 151 receptacles for the fuel assem-blies had to be welded in place with precisealignment control. This was achievedthrough use of fixturing and controlledweld sequencing. Significant distortionwas encountered in welding the top platewith its receptacles to the barrel sectionand caused major misalignment of the recep-tacles. After recovery was made throughstraightening and machining operations, agirth weld to the bottom plate assembly wasnecessary. Following significant changesand requalification of the weld procedure,a satisfactory method was developed and thegirth weld was performed without incident.

The three instrument trees, which provideoperational data for every core positionand guide tubes for the control/safety rods(Figure 2), are unique to FFTF and similarcomponents will not be used in commercialbreeder reactors. Alignment of the instru-ment trees with respect to the core andreactor head is exacting, and distortionsresulting from welding or relaxation ofweld stresses due to immersion of the treesin 1050°F sodium cannot be tolerated. The28-foot column was built up by welding aseries of seven tubular cross sectionpieces. Four of these have a non-symmetric"D" cross section and the remainder arecircular. This column was required to meeta length and straightness tolerance of+_ 0.25 and +_ 0.095 inch, respectively,in 28 feet at weld completion with noadditional machining.

After the column was completed, two boxframe weldments were attached to it, eachrequiring 36 inches of one-inch thick weld.These frames were premachined to supportthe control rod guide tubes. The box weld-ing was controlled to position the guidetubes within 0.045 inch of true positionwith respect to the column centerline.Additionally, the upper and lower frames,which control the positions of the controlrod guide tubes and which are approximately12 feet apart, were required to be orientedwithin + 0.020 inch of each other.

During the fabrication of the prototypeinstrument tree, several areas where welddistortion could occur were identified.Methods to correct these potential sourcesof distortion were evolved as the prototypewas being fabricated. These techniquesproved successful and have been incorporatedinto the fabrication of the three plant unitswhich are now in the final stages of manu-facturing. Success of these techniques washighly dependent upon the mock-ups made ofthe various configurations. Weld shrinkagewas measured so that the actual componentsfor the final product could be dimensionallycorrected to account for the expected shrink-age. In addition, local peening, fixturing,and sequential welding were employed.

The success of these measures is best demon-strated by the fact that critical measurementsmade on the prototype instrument tree beforeand af":er a 20-day exposure in 1100°F sodiumwere virtually identical. The fabricationexperience gained in developing the complexweldments has provided valuable informationto component designers faced with predictingand controlling weld distortion.

The third example of the development ofexceptional distortion control during weldingwas in the fabrication of the core supportstructure (CSS). The CSS, which is attachedto the reactor vessel by a support skirt,provides support for the core assemblies,the inner core reflectors, and the radialneutron shield which surrounds the core.In addition, it provides passages for cool-ant flow distribution to these core assembliesand to the annular space between the corebarrel and the reactor vessel wal 1. The CSSforms the pressure boundary between the highpressure inlet plenum and the low pressureoutlet plenum of the reactor vessel. TheCSS, excluding the core basket, is designedfor the 20-year life of the plant and nomaintenance is planned or envisioned.

The castings which form part of the bottomhead of the CSS are composed of two sectionsof a cone joined together by welding acrossthe ligaments of the 12 equally spaced 16inch diameter flowholes in the bottom head.The two-piece welded casting was necessitatedsince a single casting of 80,000 pounds wasbeyond the melting capacity of stainlesssteel casting suppliers. The core supportstructure required a total of six tons ofweld metal to complete its fabrication.The control of weld distortion and weldshrinkage and the production of radio-graphically acceptable welds in the heavywall sections required continuous proceduralplanning during production. Submerged arc

welding was used wherever possible and a lowdefect and repair rate was experienced. Goodaccessibility for welding was not always pre-sent due to the close proximity of parts.Tor example, two-inch thick gussets, fullpenetration welded to the inside of the bottomhead, were made with the welder standing in-side 16-inch diameter flowholes. Weldingwas performed from alternate sides of doublewelded joints to minimize distortion andinternal spiders were used in cylindricalparts to maintain roundness during girthwelding. Even with these precautions, somererounding and straightening operations wererequired.

Both the lower core support structure andthe core barrel were thermally stabilizedbefore these parts were final nachinedand joined together. This was the mostcritical weld in the core support structuredue to the tight perpendicularity tolerancesimposed. Extreme precautions were taken toinsure uniform and symmetrical drawdown ofthe barrel assembly with the lower structureby sequencing welding and monitoring relativeposition of the two structures until weldingwas completed.

With the use of 304 stainless steel as theprimary material of construction in FFTF,great care was exercised throughout manu-facture to control contamination of thefinished surfaces and to protect againststress corrosion. In several components,where dimensional stability was of concern,to maintain precise dimensions under theelevated temperature conditions, heat treat-ments were performed at temperatures inexcess of the maximum operating temperaturefor which they were designed. Most FFTFcomponents which were stress relieved uti-lized temperature-time cycles which resultedin no sensitization.

An exception to this practice was the heattreatment of the primary pump tank, shaft,and internals. These were annealed at 1150°Ffor four hours to assure that dimensionalstability was obtained at the 1050°F servicetemperature. In the case of the pump tanks,only the upper section of the pump was giventhis heat treatment. Although it was recog-nized that this heat treatment could causesevere intergranular carbide precipitation,it was thought that intergranular corrosiveattack could be controlled during manufacture.

During liquid penetrant examination of thewelds on the first primary pump tank, anunusually large concentration of very smallpenetrant indications was noted. While thegeneral size of individual indications was

such that they met Code Inspection require-ments, the large number of indications causedconsiderable concern and prompted furtherinvestigation. Information obtained by meansof surface replication and metallographicexamination of"boat samples removed from thesurface of the upper tank section revealedintergranulaf-separations having a depth froma few mils up to 100 mils. It was furtherdetermined that light surface grinding wasrequired prior to performance of the liquidpenetrant examination to reveal the indica-tions. This technique was slow and in manyinstances revealed less than 50 percent ofthe indications present, but, it was the onlymeans of assuring that the pattern of inter-granular attack could be identified. Subse-quently, eddy current inspection was substi-tuted as an alternate to liquid penetrantinspection and proved to be both reliableand efficient in determining location anddepth of intergranular attack.

It is important to note that no evidence ofintergranular attack was identified in weldmetal, in either the as-welded or annealedcondition, and in the pump shaft in spiteof the fact that the shaft had been given astabilizing anneal at 1150°F followed byapproximately 100 hours of exposure totemperatures in the sensitizing range. Forsome unknown reason, the extent of the inter-granular attack on the pump internals wassignificantly less than that observed onthe upper portion of the pump tanks, eventhough the same heat treatment was involvedand fabrication was performed in the samesupplier's shop. The pump internals werejudged acceptable and no remedial actionwas instituted. Additional restrictions,however, were instituted on the internalsto keep them clean and dry packaged (lessthan 40 percent relative humidity). Further,it was experimentally determined that nofurther intergranular attack would occur withexposure of the pump internals during speci-fication testing in water at the pumpsupplier's plant. ;

In the case of the upper tank assemblies,however, it was judged that the attack wassufficiently severe that material must eitherbe replaced or restored to a metallurgicalcondition not susceptible to this form ofattack. Following a series of heat treatscoping studies, it was determined that afour-hour heat treatment at 1800°F followedby a water quench, to be carried out withintwo minutes of furnace removal, would restorethe material to an acceptable condition. Inaddition to restoring immunity to intergrainularattack, it was desirable that the reheattreatment cycle be selected so to minimize

distoration of the final machined uppertank assembly. Prior to committing theupper tank section to this heat treatment,it was also determined, through heat treat-ment of a spare ring forging, that distor-tion during heat treatment of the pump tanksections would be acceptable. Followingthat determination, the.three upper.puniptank sections were heat treated successfullywith distortions well within acceptable limits.These heat treated upper assemblies have sincebeen rejoined to the lower tank sections, fab-rication has been completed, anri the tankshave been shipped to the FFTF site.

Efforts to determine an assignable cause forthe Intergranular attack have not providedsufficient data to establish a single agentas being responsible. Since the phenomenonwas observed on both plate and forgings, thespecific heat or form of the base materialcould not be established as being particu-larly susceptible to attack. It was demon-strated that the attack would occur in avendor's plant only on sensitized materialand that an Incubation period of approximatelyfive weeks was required to produce detectableattack.

While the remedial actions taken to restoreimmunity of the tank sections to intergranularattack were successful, they were most unde-sirable in terms of schedule and componentcosts. This experience again emphasizes thenecessity to thoroughly examine the implica-tion of any stress relief of unstabilizedaustenitic stainless steels in the sensi-tizing range of 900-1650°F. If such ananneal is found to be necessary, elaborateprecautions to control cleanliness and mois-ture are mandatory through the fabrication,storage and erection operations.

REACTOR VESSEL

The reactor vessel is constructed of Type304 stainless steel, and consists of fourbasic sections: the barrel, core supportring forging, inlet plenum and liner. Thebarrel, which comprises the upper cylindri-cal portion of the vessel, has an insidediameter of 20 feet, 3 inches, and a wallthickness of 2 3/8 inches. The barrel isjoined to a ring forging located at aboutcore midplane, which provides support forthe core support structure. The ringforging is joined to a torispherical bottomclsoure which serves as a high pressureInlet plenum for the reactor coolant. Theoverall height of the reactor vessel is43 feet, 1 1/2 inches.

The fabrication and in-shop handling of thereactor vessel was complicated by its sheersize and its large diameter to wall thicknessratio which resulted in significant distortionunder its own weight when lying horizontally.Special fixturing was used to control thisdistrotion for some final machining operationswhile in other cases, the machining operationfactored in and compensated for the distor-tion. Distortion control in the welding ofthe core support structure and the thermalliner into the reactor vessel required theuse of controlled weld bead deposition andcontinuous alignment checks during thewelding.

Because of the low temperature, low exapnsioncarbon steel reactor vessel head and the hightemperature, high expansion stainless steelvessel, a typical bolting attachment wouldresult in ratcheting of the vessel wall.Therefore, a specialized arrangement wasutilized. A ring forging is welded into thetop of the ban-el directly below the top flangeto which 30 support arms are welded. Thereactor vessel it securely attached to themain support structure (MSS) by means of bolt-ing the vessel support arms (Figure 3) to theMSS. Sealing is accomplished by an omega sealwelded to the upper flange of the reactor vesseland to the I-ring and by two concentric metallic0-rings between the head and the I-ring. Thelong retention bolts are preloaded only in themiddle to achieve sealing loads and stillallow the head to move freely with respectto the reactor vessel support ring. Thesupport arms remove the mechanical loads fromthe high thermal stress protion of the vesselwall. The vessel support arms are also slotted,which eliminate any stress due to the tempera-ture gradient, but more important acts as aresilient spring to reduce loads on the con-crete ledga.

The reactor vessel will receive an ASME CodeN stamp in the field following the weldingof the omega seal to the reactor vesselflange and helium leak testing in accordancewith ASME Code Case 1595. Because reactorvessels for the future LMFBR plants will notbe that much largerthan "the FFTF vessel andhead, the experience gained in the design,fabrication, and installation of the FFTFvessel provides the technology needed forthese future vessels.

MAIN COOLANT PUMPS

Prior to FFTF, the stage of development forliquid metal pump design in the U.S. wasreflected in EBR-II and Fermi-I. The majorcharacteristics of the sodium pumps used inthose facilities and FFTF are presented inTable I. It is clear that the FFTF pumpmust satisfy significantly more demandingrequirements than its predecessors; conse-quently, a definite advancement in liquidmetal pump technology will be achieved withthe FFTF design.

The pump (Figure 4) is a free surface, shaftseal, single stage, single suction pump withat variable speed drive and is located in thehot leg of the reactor heat transport system.Inlet flow enters the hydraulic assembly viathe suction elbow and is discharged from theimpeller radially into the diffuser/turningvane assembly which in turn redirects theflow from a radial to a downward directionthrough the discharge nozzle.

The 1O5OCF design temperature and thermaltransient requirements for the pump neces-sitated in some cases the attenuation oftransients by mixing the transient fluidwith constant temperature fluid. Thehydraulic assembly represents a unique

TABLE 1 U.S. LIQUID (METAL PUMP DESIGN COMPARISONS

CHARACTERISTIC

PUMP TYPE

PUMP HEAD (Ft Na)PUMP CAPACITY (GPM)OPERATING TEMP (°F>APPROXIMATE BHP

FACILITY

EBRD

MECHANICAL-CENTRIFUGAL

1354,500

700500

FERMI 1

MECHANICAL-CENTRIFUGAL

32011,800

600*1,000

FFTF

MECHANICAL-CENTRIFUGAL

50014,5001,0502,500

'Designed For A Maximum Temperature Of 1000°F

approach to this problem, in that only non-pressure containing walls are exposed to thefull transient. Strain ratcheting is practi-cally eliminated by this approach.

The pump shaft is fabricated by welding twosolid forgings to a tube section. The dyna-mic design criteria imposed in conjunction:with the fact that only two radial bearingsare used required the use of an 18 footi28 inch diameter section madeifrom)a=seam-less extruded tube, A thermalIbaffleTsys-tem is located in the upper iendof" the tubeto minimize the heat transferred into theshaft seal region.

Briefly, some of the pump development programscarried out or still in progress are:

1. Design verification of the seal in aconventional test rig and testing ofd complete shaft seal package in asimulated operating environment withseal operation over a sodium pool, inargon cover gas.

2. To confirm bearing performance calcula-tions, tests were performed in a watertest rig followed by bearing operationaltests in sodium.

3. Bearing hardfacing material selectionevolved from a series of thermal shocktests of coupons hardfaced with candi-date materials. Stellite proved superiorwith respect to tolerating repetitivethermal shocks. A follow-up to this workconsisted of thermal shock tests on full-size hard-faced journal and bearings.

4. An air model test of the inlet configura-tion was conducted which confirmed thecalculated fluid velocity profilas atthe impeller suction.

5. Two development programs aimed at support-ing the dynamic design effort were per-formed. One test provided data en theeffects of shaft motion transmitted tothe pump structural members via thesurrounding fluid to enable accurateanalytical modeling. The other testexposed a full-size shaft to simulatedplant thermal gradients for evaluationof thermal distortions. As a result,design modifications were made and theshaft satisfactorily completed retesting.

6. A comprehensive prototype pump testprogram in water followed by sodiumtesting at the Liquid Metal EngineeringCenter is planned prior to installationof this equipment in the FFTF.

The primary purpose of the prototype isto demonstrate performance characteristicsand mechanical reliability at operatingtemperatures prior to actual in-plant use.At the completion of the scheduled test-ing, the prototype is potentially capableof serving as a spare for the FFTF plant.

The water test of the prototype pump has beencompleted, and it has verified the basic mec-hanical design and hydraulic performance char-acteristics. ̂ During performance of this phaseof the prototype pump test, two problems wereidentified for which design changes wererequired. A shaft orbit was observed as mea-sured by journal position sensing instrumen-tation in the hydrostatic bearing. The shaftorbit has been reduced to an acceptable levelby a reduction in the hydrostatic bearingsupply orifice size and by an improved con-centricity between the shaft and its surround-ing chamber. A slight dip in the head flowcurve was detected and attributed to flowseparation in the diffuser. This conditionwas also brought Within acceptable limits bya modification to the impeller dischargeconfiguration.

The first FFTF plant pump was modified toincorporate the changes made to the prototypepump and is presently being tested in water.Over 300 test hours, to date, have verifiedthuC the modifications made for the two prob-lems identified during the prototype pumptests were acceptable.

Testing of the prototype pump in high tem-perature sodium is scheduled to start inJuly 1975. This testing represents a firstin that the pump will be operating at. the1050°F temperature level for extended periods,Basictesting will be essentially the same asthat performed in the water facility, withthe additional capability of subjecting thepump to thermal transients approximating thoseimposed by the FFTF plant.

THE INTERMEDIATE HEAT EXCHANGERS

Three intermediate heat exchangers (IHX) of133 cWt are used in FFTF. The units arevertical, counterflow, shell and tube typeheat exchangers. The IHX is composed ofthree main subassemblies: the hangingsupport, which is a cylinder bolted to thecontainment building operating floor andsupports the other two subassemblies, theshell, and tube bundle. The IHX has anoverall length of approximately 35 feet, aninside diameter of 76 inches, and has anempty weight of 90 tons.

In contrast to heat exchangers used in lightwater reactor systems, the primary sodium ison the shell side and the secondary sodiumcirculates through the heat transfer tubes.Because the tubesheets are welded to a rigidcentral downcomer pipe, provision was madefor the differential thermal expansionbetween the tubes and the cooler dbwhcoiTierpipe. This was accomplished by adding alarge compound bend in each tube (Figure5). The IHX for the Clinch River BreederReactor Plant (CRBRP) will use a floatinglower tubesheet with a bellows expansionjoint in the central downcomer to accom-modate the differential thermal expansionand will, therefore, use straight heat trans-fer tubes.

The manufacture of the heat exchangers wasaccomplished without an unusual number ofdelays and problems, considering the com-plexity of the design and the severalthousand individual parts involved. Theextensive use of well designed jigs andfixtures minimized dimensional deviationsand weld distortions and permitted inter-changeability of parts between units.

One fabrication operation for which littleproduction experience existed was the tubeto tubesheet spigot welding. To eliminatethe crevice between the tube and tubesheethole, which exists when a conventional frontface tube-to-tubesheet weld is used, thisdesign employs a full penetration fusionbutt weld of the tube to a machined spigoton the back side of the tubesheet. An auto-matic tungsten inert gas torch fuses thejoint from the inside of the tube. The tube-sheet downcomer assembly was erected verti-cally in the clean room and tubes were weldedinto both tubesheets simultaneously. Allwelds were accepted by a visual as well asa radiographic inspection before the nextrow of tubes were fitted. Host of the welddefects found were porosity and 'linearindications due to sharp changes in weldcontour which were corrected by refusingthe weld.

While welding the eight row of tubing on thelead tube bundle, it was noted that some weldswere not fusing smoothly at the outer cornerof the spigot and the corner produced a linearindication on the radiographs. Visual exami-nation disclosed an unusual oxide coating onthe weld which upon chemical analysis wasdetermined to be aluminum oxide. Furtherchemical analyses of the tubing heats involvedin the oxide coated welds revealed an aluminumcontent which varied from 0.005 to 0.40 per-cent. Vesting determined aluminum content

up to 0.15 percent does not degrade thestrength of 304 stainless steel. However,an aluminum content above 0.07 percent doesincrease the fluidity of the molten poolduring the fusion welding, producing sharpchanges in weld contour and an oxide coatingwhich could possibly crack off and be carriedinto the pump bearings. All tubes containingaluminum in excess of 0.07 percent were re-placed with other tubes. While FFTF did notspecify an aluminum content for heat transfertubing, the specification for the CRBRP IHXincorporated a 0.07 percent maximum aluminumcontent.

Major testing programs performed to confirmoperating and design features of the IHX arebriefly described below:

1. A seven-tube model was tested with waterto measure pressure loss characteristicsthrough the tube support plate flowhoiesand to measure tubeside flow inducedvibration at five different flow com-binations. No vibration occurred from10 to 150 percent of full flow.

2. A single-tube vibration test was per-formed which verified that tubeside flowdoes not. induce significant vibrationsin the bent tube section and that sup-port was not required at the center ofthe bend.

3. A tube fretting wear test was performedin 1050°F sodium utilizing a three-tube,two-span mockup of the tube bundle inwhich flow induced vibrations were simu-lated. It was found that the frettingwear, for up to 200 million cycles, wastoo small to express in quantitativeterms.

4. The IHX employs a perforated and slottedconical flow distribution plate at thetop of the primary side inlet plenum toevenly distribute the flow to the shellside of the tube bundle. In order todetermine the hole and slot pattern whichwould provide even flow distributionaround the periphery of the tube bundle,a one-fifth scale model of the inletplenum chamber and inlet elbow was builtof plastic and tested in a water testloop. The model was divided into twelveequal sections. By varying the totalhole area in each section, an equalizedflow pattern was determined., Flow ratesup to the equivalent of 100 percent ofdesign flow were used and a satisfactoryflow balance was establishecL The finalconfiguration flow balance is +_ 2.5

percent down to 60 percent of full flowand +19 and -6 percent at 40 percentflow.

5. A full-scale model covering the heattransfer section between the tubesheetswas built. The model was a 72degree^segment of the heat exchanger crosssection. The model was used to deter-mine the velocity distribution on theprimary side between the tube supportplates from the center to the outside ofthe tube bundle and to measure the pres-sure drop across each support plate.Measurements were taken in the range of25 to 100 percent of design flow. Thetests revealed that adjustments warenecessary in the number and size offlow holes in some regions. Modifica-tions were made and tested to achieve amore constant velocity distributionthrough each support plate. The pres-sure drop measurements taken after themodifications confirmed that the pres-sure drop on the primary side would meetspecification requirements.

Following the above testing, the model wasreduced to a 48 degree wedge section topermit flows up to 150 percent of thedesign flow rate. Testing was performedto determine the amplitude and frequencyof flow induced vibrations. It was shownthat no resonance occurred between thedriving frequency and the tube naturalfrequencies. The tube response was lessthan 0.001 inch rms and typically randomin nature.

6. In order to verify acceptable flow distri-bution into the secondary side, a one-thirdscale model was built which simulated thecenter secondary sodium downcomer liner,the lower hemispherical head, and the tubebundle. The model was tested without anyflow distribution structure in the headand unacceptable flow balances resulted.Testi ng revealed that an annular ri ng,attached to the inside wall of the head,would produce the desired distribution.

7. The ability of the tube-to-tubesheet buttwelds to withstand imposed thermal shockswas demonstrated by means of a seven-tube,double tubesheet, stainless steel tiltingautoclave-type vessel using liquid sodiumas the heat transfer medium. A total of252 thermal shocks, using the most severetemperature changes the welds will exper-ience, were performed by heating the tube-sheets and sodium to the two temperaturesrequired to produce the desired differentialtemperature and then rotating the autoclave180 degrees such that the sodium ran throughthe tubesheets and tubes to the opposite

8

end of the autoclave. Following theshock testing, the 14 tube-to-tubesheetwe'ds were subjected to inspections andexamination and no evidence was found ofcracking or separation of tubesheetcladding from the base metal.

PROTOTYPE TESTING

To confirm design adequacy of the more develop-mental reactor and plant components, a numberof prototypes were built and are being testedat operating temperatures. Included in theprototype test program are the primary sodiumpump, in-vessel handling machine (IVHM), corerestraint mechanism (CRM), instrument tree(IT), and control rod drive mechanism (CRDM).At this point in time, the primary sodium pumphas completed water testing and is awaitingtesting in sodium, the CRM has satisfactorilycompleted air testing and is currently insodium testing, and the CRDM has satisfactorilycompleted a life test in 1TQO°F sodium. Testrequirements for the IVHM and IT are brieflydescribed below.

The incorporation of the special in-reactortesting components imposed the developmentof IVHM and IT designs unique to FFTF. Thesecomponents are designed for a sodium environ-ment of 1100°F at one end and an ambientenvironment 30 feet away at the control end.The components remain static during reactoroperation but are required to move with pre-cision during shutdown. These components havebearings immersed in sodium and bearings andseals exposed to sodium vapors which cancause a sodium frost problem. These com-ponents have sliding surfaces with closetolerances subject to sodium frost formationwhich must continue to slide freely in orderto adequately perform their function,_ Thetest program was specifically focused ondetermining the retention of these operatingcapabilities throughout design lifetime.

In the case of the IVHM (Figure 6), thedesign task was to make a 50 ton machinemounted on an adjustable foundation, index94 different remote locations over an areaof approximately 104 square feet and be ableto locate its grapple within +_ 1/4 inch.It must maintain capability to lift thecomponent from the core and position it,either in another core location, or in astorage module for later use or examination.During these operations, the arc made bythe grapple is dependent on accurately con-trolling the rotating plug and rotating armto insure that it does not contact otherreactor components. All of these operationsare performed without visual aid under 16feet of sodium and must be done accuratelyover the lifetime of the component.

Although operating requirements for the ITare not as stringent as for the IVHM, a dis-connect must be made of the CRDM drivelineprior to each refueling, the in-reactorassembly must be moved to a parked position;af_te.njre.fueling,_the. IT is_ moved in place6ver'*the icore?and fa;s«cohnecti on -muTf * bestiiadeof the CRDMSdriyeTine.

The prototype IT and IVHM are presently under-going testing at Hanford Engineering Develop-ment Laboratory. This tast program consistsof two miljor phases: (1) air testing in thecore mechanical mocfcup (CMM); and (2) hightemperature sodium testing in the CompositeReactor Component Test Activity (CRCTA).

Each program consists of individual and com-posite, functional, dimensional and interfacetests with a simulated core and other reactorcomponents. In CRCTA these components areimmersed in sodium and are exposed to thethermal environment that the plant units willbe subjected to during plant operation.

The IT has successfully completed functionaltesting in CMM and initial functional anddimensional tests in CRCTA. The crucialphase of CRCTA testing occurred after thetree was soaked in 1100°F sodium for 20 days.Prime concern centered around the potentialrelaxation of welds and resulting stressrelief which could cause in-reactor distor-tions making the tree inoperable. Functionaltesting of the IT following the 1100°F soakwas successfully completed and dimensionalmeasurements before and after the 20 dayexposure to 1100°F sodium indicate relativelylittle or no distortion of the in-reactorassembly.

Air testing of the IVHM prototype revealedthat extensive changes in design detailswere necessary to correct problems whichwere not evident on paper. These changeshave been incorporated into the plant unitcomponents. The IVHM has successfully com-pleted air testing in CMM and is currentlybeing installed in CRCTA for composite testwith the IT and other reactor components.

All the test programs, models, mockups, pro-totypes, etc. are too numerous to discusshere. But the significant accomplishment ofthe test programs is that the testing todate has confirmed the adequacy of the designsestablished by the codes,/standards, andanalytical methods developed to produceFFTF components.

SUMMARY

Just as FFTF has prototype components toconfirm their design, FFTF is serving as aprototype for the design of the commercialLMFBR's. The operating conditions of LMFBR's,i n-parti cul ai?-the -eleyajted rtemperf ture-condi-ti ons, ireciui red an extensi on, of:. the,. state-of-the-art pf jnechanical design which has been,accoipTishecl^by"FFTF. Development of the FFTFdesign to meet stringent safety and performancerequirements necessitated the extension ofmany analytical"toolssneeded for LMFBR's.

Design and maniifiictufe of- critical componentsfor the FFTF sysfern[ h3ve-|een afccomsnsnsdprimarily using vendors with little or noprevious experience in supplying componentsfor high temperature soilurn*systems. Theexposure of these suppliers, and through thema multitude of subcontractors9 to the require-ments of this program has been a necessaryand significant step in preparing Americanindustry for the task of supplying the largemechanical components required for commercialLMFBR's.

In conclusion, FFTF is a vital, necessarystep in the United States program for thedevelopment of an economical, reliable, andsafe commercial LMFBR. The FFTF has currentlycompleted many of its originally establishedgoals and accomplishment of the remainderwill provide further significant informationnecessary to the LMFBR development.

REFERENCES

1. M. Shaw, "The Role of FFTF in the U.S.LMFBR Programme," Nuclear EngineeringInternational, Volume 17, pages 613-628, published by I PC Electrical-Electronic Press Ltd., London, August1972.

2. J. E. Nolan, J. J. Morabito, A. A.Simmons, and D. J. Cockeram, "FastFlux Test Facility," Proceedings ofthe International Conference Organizedby the British Nuclear Energy Societyheld on 11-14 March 1974 at the Insti-tution of Civil Engineers, London,England, pages 71 to 84, published byThomas Tel ford Limited, London.

Figure 1. Core Basket Upended During Fabrication

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MAST AND <

SUPPORT ASSEMBLY

CONTROL ROD

GUIDE TUBE

INSTRUMENT

ASSEMBLY

GUIDE TUBES

„ PLUG

ASSEMBLY

IN REACTOR

ASSEMBLY

Figure 2. Instrument Tree

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Figure 3. Reactor Vessel/Closure Head and Main Support Structure Interface

12

SHAFT SEALASSEMBLY

FLOOR LINE

420.5 IN.

SHIELD PLUG

SODIUM LEVELINDICATOR WELL

THERMAL 3AFFLES

MAXIMUM SODIUM LEVEL

UPPORT CVLINDER

SHAFT

SODIUM DISPLACEMENTCHAMBER

MINIMUM OPERATINGSODIUM LEVEL

BEARING SUPPLYRESERVOIR

IMPELLER

OIFFUSER TURNINGVANE ASSEMBLY

PUMP TANK

Figure 4. Primary Sodium Pump

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Figure 5. IHX Tube Bundle During Tube-to-Tubesheet Welding

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REACTOR REFUELING [•PLUG (RRP) 1

RIGID ARM

GRAPPLE ARMGUIDE ROLLERS

GRAPPLE CAMROLLER

GRAPPLE

CORECOMPONENT

HOLD DOWNTOE

DRIVEASSEMBLY

TOP OF REACTORVESSEL HEAD

STEM/GRAPPLE ARM

STEM LOWERSUPPORT ROLLERS

STEM ROLLERTRACKS

Figure 6. In-Vessel Handling Machine

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