Text

Letter from J. G. Herbein, Meted/Gpu, to Denwood Ross, USNRC, Transmitting Safety Analysis Report for Transition to Natural Circulation
	ML060760219
Person / Time
Site:	Crane
Issue date:	04/12/1979
From:	Herbein J; Metropolitan Edison Co
To:	Ross D; NRC/FSME
References
	FOIA/PA-2006-0120, GQL 0509
	Download: ML060760219 (211)
	v • d • e

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r i Metropolitan Edison Company-lli, ois _.Post Office Box 480 Middletown.

Pennsylvania 17057 .717 944.4041 April 12, 1979 GQL 0509 Mr. Denwood Ross Assistant-Director. Division of Reactor Safety U. S. Nuclear Regulatory Commission D.C. 20555 I.

Dear Sir:

Three Mile Island -Nuclear Station Unit 2 (TMI-2)-License No. DPR-73 Docket No. 50-320 Safety Analysis Report for Transition to Natural Circulation (C-D)Enclosed please find the Safety Analysis Report and preliminary information for the proposed transition to long term natural circulation at TMI-2, as requested at the Commission Meeting of April 9, 1979. This report is current as of approximately April 9.The analysis is significant in that it suggests a great deal of flexibility in placing the TMI-2 reactor in a natural circulation mode. Additional analyses -are continuing to be performed which-wil l-fine more specifically the .proposed final end point tempera-ture and pressure conditions as well as state points in the various supporting plant. systems. Detail analyses and procedures on the exact methods for achieving natural circulation are also in work.As additional information becomes available, we will supplement the attachment as necessary. It is our conclusion, from the data in the attachment, that long term natural circulation is a viable way for placing the TMI-2 *reactor into a long term stable condition, and the safest of the various options available. Si cere3~y, -.Original signed/ *B*A J. C. Herbein Vj. G. Herbein .Vice President Generation

JGH:LIWJH:al Enclosure

..* cc: Harley Silver (NRC)-::ro po'n Ed.so.1 Cornp.ny is a Mrnbacr of ihe Gen'cral PLRbc Uti.1lir'- qwt-- CORE THERMAL BEHAVIOR 79aOrI900ro CONTENTS VPv 1) "3.7 CORE THERMAL BEHAVIOR", G. A. MEYER/A. B.. JACKSON, APRIL 10, 1979.2) "CRITERIA DURING ESTABLISHMENT OF NATURAL CIRCULATION," L. L. LOSH/J. F. BURROW, APRIL 10, 1979.-3) "RELIABILITY AND UNCERTAINTY OF THERMOCOUPLES FOLLOWING LOSS OF FEEDWATER TRANSIENT," T. L. WILSON, APRIL 5, 1979.4) "ACTION ITEM 143," J. T. WILLSE, APRIL 6, 1979.5) "RESPONSE TO THERMOCOUPLE REQUEST," J. A. WEINER, APRIL 5, 1979.6) "REQUIRED FLOW FOR CORE COOLING," A. B. JACKSON, APRIL 10, 1979.7) "COOLDOWN PRESSURE," J. R. GLOUDEMANS, APRIL 10, 1979.8) "ADIABATIC HEATUP RATES," J. H. JONES, APRIL 10, 1979.9) "INCREASED T.C. READINGS DUE TO PROXIMITY OF FUEL PARTICULATES," P. J. HENNINGSON, APRIL 10, 1979.10) "BOILING CONDITIONS IN CORE," J. A. WEIMER/R. L. HARNE," APRIL 1, 1979..11) "MINIMUM CORE FLOW -LONG TERM COOLING," G. A. MEYER, APRIL 4, 1979.12) "CORE FLOW DISTRIBUTIN FOR ONE PUMP AND TWO PUMP OPERATION," R. M. HIATT, APRIL 10, 1979.13) "CORE BYPASS FLOW FOR CORE BLOCKED AT TOP ONLY," R. M. GRIBBLE, APRIL 8, 1979.14) "INCORE THERMOCOUPLE ERROR EVALUATION," J. A. WEIMER, APRIL 10, 1979.15) "DISCREPANCY BETWEEN THERMOCOUPLES AND OUTLET RTD TEMPERATURE MEASUREMENTS," T. L. WILSON, APRIL 9, 1979.16) "LiYNXi MODEL FOR TMI-2 BLOCKAGE STUDY," R. M. HIATT, APRIL 10, 1979.17) "DAMAGE MODEL -FLUIDIZED BED," P. J. HENNINGSON, APRIL 10, 1979.18) "ESTIMATE OF LOOSE CORE DEB3RIS VOLUME (4/9,'79 -2000),"CORE CONDITION TASK FORCE, APRIL 9, 1979.! 1) .c~ BaspEg~uf E~i ,,s J4 gS C,> L 4 %'OAI C c ALC(LA-/Or J 5 B4 P- L1. J 4Apse} q 2 /l977 20) CONTENTS 3. SAFETY EVALUATION INFORMATION FOR TRANlSITION TO 'NATURAL CIRCULATION, COOLING.3. 1 DESCRIPTION OF COOLING MODE 3.2' 'CONTINGENCY PROCEDURES 3.3 -SYSTEM PERFORMANWCE'ANALYSIS -IN -NATURAL CIRCULATION .3.4 CECKPOINTS DURING.THE'TRANTSITION OPERATO ONTRLCRUAIN 3.5 -HY'DROGEN OAS-CONSIDERATIONS-3 .6 PRESSURE TEMPERATURE CONSIDERATIONS 3'.7 CORE 1MELT'CONSTIDERATIONS 3.8- CORE THERMAL BEHAVIOR.3.9 CRITICALITY, CONSIDERATIONS 3.0 SAFETY EVALUATION INFORMATION-FOR TRANSITION TO NATURAL-:CIRCULATION COOLING.METROPOLITAN EDISON CO. HAS EVALUATED' THE VARIOUS STATES FOR MAINTAINING THE TMI 2 REACTOR A LONG TERM COOLING MODE. WE HAVE PREPARED THE FOLLOWING EVALUATIONS WHICH DEMONSTRATE.THAT THE REACTOR AND ASSOCIATED SYSTEMS CAN -SAFELY UTILIZE RCS NATURAL CIRCULATION CORE COOLING WITH THE STEAM GENERATOR:SECONDARY SIDE IN A SOLID FLOWING.1WATER CONDITIONS FOR HEAT REMOVAL. -.......-' ' ' ... : :.............. ..... ...:.-......-.-: .: y : ...R -... .........................* ..-'., :....': :. : .,.......: .., ..:...'. .!,.........' ' ' ...... .: .. ...........: .. ......... j; ' ...: ' '-"' :.':......* ..: :....'. ., ', , ' .... .

3.1 DESCRIPTION

OF. COOLING MODE ATTACHMENT i TO THIS REPORT ENTITLED "A

SUMMARY

OF"NATURAL CIRCULATION-ALTERNATIVES FOR LONG-TERM CORE COOLING AT TMI-2 "DESCRIBES THE PROPOSED METHOD FOR LONG TERM COOLING. TTHE DOCUMENT CONTAINS DETAILED INFORMATION ON THE RECOMMENDED COOLING METHOD ALONG WITH DISCUSSIONS DOCUMENTING THE SUPERIORITY OF THE RECOMMENDED METHOD OVER ALTERNATIVE CONSIDERATIONS.- I .....v I..I..m .: , m ., -m I 4... .3.2 BACK.. .. ETR UP CONSIDERATIONS. OPOLITAN EDISON CO. HAS PREPARED DETAILED NT OF EQUIPMENT MALFUNCTION-A -LIST. OF TH ENT 2. THE BACKUP CONTINGENCIES'PROVIDED-ACCURATE RESPONSE TO.EllERGENCY OR OFF NORN IN THE EVE'IN ATTACHID RAPID AND.I ....OPERATING INSTRUCTIONS ESE PROCEDURES IS PROVIDED BY THESE PROCEDURES ASSURES AL- PLANT CONDITIONS..

3.3 SYSTEM PERFORXANCE ANALYSIS IN' NATURAL-CIRCULATION

.B&W HAS PERFORE DEALDAAYE-FTENATURAL CIRCULATION CONDITION FOR THE RECOMM1ENDED NODE OF COOLING. THESE CALCULATIONS INCLUDE BOTH HANTD CALCULATIONS AND THE.DEVELOPMENT CIF COMPUTER CODES TO PREDICT TRANSITION SYSTEM RESPOINSES.. THE ANALYTICAL -TECHNIQUES USED ARE'DESCRIBED'IN ATTAC1HMENT 3.- THESE TECHNIQUES HAVE BEEN BENCID4ARKED AGAINST.NATURAL CIRCULATION DATA OBTAINED AT DAVIS-BESSE

1. IN ADDITION, TESTS 'WERE PERFORMED AT B&WS ALLIANCE RESEARCH'* CENTER WHMICH DEMONSTRATETHE-EXCELLENT COOLING CAPABILITIES.USING THE OTSG"S IN THE RECOMMENDED COOLING MNLODE. .THE.RESULTS OF THEEALLIANCE TESTING .ARE'DISCUSSED-;IN DETAIL IN ATTACHMENT 1..-ATTACHMENTS 1, AND 4 DISCUSS THE POTENTIAL FOR CORE BLOCKAGE.-

AS SHOWN IN FIGURE 2 OF ATTACHIET1 -OETHNADEQUATE CO-RE FLOW WILL EXIST FOR THE-RANGE OF ESTIMA.TED.BLOCKAGE. IN ADDITION ACCEPTANCE CRITERIA HAVE BEEN PREPARED WHICH WILL'BE USED TO SAFELY TERMINATE' THE TRANSITIONi TO THE NATURAL-CIRCLATON NDE F:NEEDED.IT IS THEREFORE METROPOLITAN. EDISON: COMPANY'S. VIEW THAT.DET.AILED'ANALYSES OF THE TRANSITION'TO THE NATURAL CIRCULATION MODE OF COOLING DEMONSTRATETHAT THE RECOMMENDED COOLING MODE CAN 'MAINTAIN THE -'CORE' IN -A'SAr3E .CON~DITION.. IN ADDITION, IN THE UNLIKELY EVENT THAT 'PROBLEMS DO ARISE, ACCEPTANCE CRITERIA WILL ASSURE THAT THE TRANSITION OPERATION CAN 'BE- SAFELY' TERMINATED. AND% THE- PLANT RETURNED TOITS ORIGINAL:COOLING MODE. % '3 .4 CHECKPOINTS DURING THE TRANSITION OPERATION TO NATURAL CIRCULATION. .-THE ACCETANCE CRITERIA FOR THE TRNASITION OPERATION ARE INCLUDED IN ATTACHMENT 1 AND- THE THERMOCOUPLE CRITERIA ARE INCLUDED IN ATTACHMENT 4.I --..... : m ....I...m , :.i .: . 3.5 .HYDROGEN EVALUATION METAL-WATER REACTION D'CRI G THE INITIAL PHASES OF THE TMI-2 INCIDENT GENERATED LARGE QUANTITIES OF HYDROGEN ON MARCH 28,:1979. THIS HYDROGEN FORMED A BUBBLE WHICH BECAME TRAPPED IIN THE HEADOF THE REACTOR VESSEL.' THE.PARTIAL PRESSURE-OF HYDROGEN IS THIS.BUBBLE CAUSED THE REACTOR COOLANT TO'BECOME SATURATED:WITH HYDROGEN. AFTER THE BULK' OF.THE BUBBLE WAS REMOVED ON APRIL 1, THE COOLANT REMAINED SATURATED WITH 1300 TO 1400 STDh CC OF HYDROGEN PER KILO-GRAM OF COOLANT. EXTENSIVE -DEGASSING OF THE REACTOR COOLANT DURING THE 'TINE PERIOD FROM APRIL 2 THROUGH APRIL 8 IS BELIEVED.TO HAVE SIGNIFICANTLY REDUCED THE CONCENTRATION OF -DISSOLNVEDwh-YDROGEN. HOWEVER, SOME HYDROGEN GAS.WAS.BELIEVED .TO HAVE BEEN TRAPPED IN THE .CONTROL ROD DRIVE MECHANISMS (CRDMs) AND HAS-NOT..READILY DISSOLVED:INTO'THE RE:CTOR COOLANT. SO, ON APRIL 9TH, THE REACTOR.COOLANT SYSTEM PRESSURE WAS CYCLED TO PROGRESSIVELY LOWER PRESSURES, REACHING A MINIMUM PRESSURE OF'1411 PSIG. THIS EXPANDED'TEE GAS TRAPPED IN THE CONTROL ROD'DRIVES AND ALLOWED IT TO BE ENTRAINED IN THE RC.FLOW7. THE AC NOISE SIGNALS ON THE REACTOR COOLANT'PRESSURE TRANSMITTER CONFIRMED.THAT BUBBLES WERE..RELEASED .EACH TIME THE PRESSURE.REACHED A NEW LOW. (BUBBLES APPARENTLY ALTERNATE THE NOISE SIGNAL -AND REDUCE THE -PEAR-TO-PEAK FLUCTUATION).

THEREFORE, IT IS CLEAR.THAT, AT PRESSURES.

ABOVE 411 PSIG, THE GAS WILL:BE COMPRESSED FAR BACK INTO THE CCRDMs AND THAT. THE REACTOR COOLANT SATURATION PRESSURE IS BELOW 411 PSIG. : ' SINCE THE SOLUBILITY OF HYDROGENN.WILL DECREASE AS THE TEMPERATURE DECREASES, NATURAL -CIRCULATION MUST BE PERFORMED AT.A PRESSURE SUFFICIENTLY ABOVE 411 PSIG TO ASSURE THAT ANY DECREASE IN SOLUBILITY DUE TO TEMPERATURE IS OFFSET.BY.THE SOLUBILITY. iNCREASE 'DUE TO PRESSURE. IF THE IfINIMUM TEMPERATURE EXPECTED'DURING NATURAL CIRCULATION.IS .140 0 F, AN OPERATING PRESSURE OF 600 PSIG OR-'GREATER WILL ASSURE THAT NO BUBBLES) ARE FORMED,EVEN IF IT IS ASSUMED THAT REACTOR'COOLANT IS PRESENTLY SATURATED AT 411 PSIG. ACTUALLY, THE REACTOR COOLANTS'HYDROGEN SATURATION PRESSURE IS-. EXPECTED TO BE SIGNIFICANTLY BELOW 411 PSIG, BUT THIS WILL NOT BE ABLE TO BE PROVEN BY PRESSURE.REDUCTIONS DUE TO NPSH LIMITATIONS'ON THE RC PUMS. IN ORDER TO DETERMINE THE.ACTUAL SATURATION LI-MIT, PRESSURIZED REACTOR.COOLANT SA4PESWILHAVE TO BE ANALYZED FOR DISSOLVED HYDROGEN THE NET PRODUCTION OF RADIOLYTIG IJY]ROGEN1 OR OXYGEN IS EXPECTED TO BE ZERO. AS LONG AS THE PARTIAL PRESSURE.OF HYDROGEN IN THE REACTOR COOLANT SYSTEM IS KEPT IN THE RANGE OF 5 TO 15 PSI,(REF. -1).ADDITIONA EQUIPMENT ISNEE OASR DQUATE RCS.DEGASSING CAPABILITY TO REMOVE -ENOUGHGAS .FROM'THE-SYSTEMITO ASSURE'EVENTUAL-DEPRES9SURIZA-. TION FROM 600.I'SIG.TO AnTMOSPHER.IC PRESSURE.WITHOUT INTERRUPTING COOLANT FLOW.REFERENCES' 1.WATER COLN'EHOOYOFPWRRATRBY PAUL 'COHEN, GORDON ANDV BREACH. SCIENCE PUBLISHERS OF NEW YORK, 1969

3.6 PRESSURE-TEMPERATURE CONSIDERATIONS
FERENCES:

1) J.H. TAYLOR TO DISTRIBUTION,~

SANE 'SUBJECT, .4/9/79,' 8:53 -P.M.2) C.E. HARRIS TO -C.W. -PRYOR, "P.T.. LIMITS 'FOR LONG TERNX COOLING," 4/10179, 5:50 P. M. /IN RESPONSE TO REFERENCE -1) , THE FOLLOWIN.G.STATZE~NT IS PROVIDED AS INPUT TO: SECTION '3.8'OF THE SUBJECT SER. THIS INPUT IS BASED'0N:THE ANALYSIS RESULTS DOCUMENTED IN .REFERENCE 2).'BASED ON FRA:CTURE MECHANICS ANALYSES OF THE TMI-2 REACTOR VESSEL, P~RESSURE~--TEMPERATURE LIMITS FOR'LONG TERN: -COOLING OPERATION HAVE BEEN ESTABLISHED. THE ANALYSES WERE CONDUCTED'IN ACCORDANCE WITH APPENDIX:G TO SECTION.III OF,-A S 1-E CODE.FOR' ACCIDENT CONDITIONS.' .THE'.CALCULATIONS ARE -APPLICABLE..TOR-FLAW--DEPTHS UPTO:ONE QUARTER OF THE REACTOR VESSEL THICKNESS'ADSOL ONEVTVL BOUND ANY FLAWTS WHICH MIGHT.EXIST IN SERVICE. -`THESE ANALYSES -CONSIDERED'A WORST CASE TRANSIENT-ASSOCIATED WITH. HPI SYSTEM OPERATION BY. CONSERVATIVELY. ASSUMfING THAT NO MIXING OF HPI AND REACTOR COOLANT.* WATER:OCCURS IN.THE INLET.PIPING. FOR THIS: CASE, 'THE: REACTOR: VESSEL INLET'NOZZLE IS THE GOVERNING-WELD;.A PLOT OF THE ALLOWABLE PRESSURE!-TEMIPERATURE ENlVELOPE SATCHD TESYTE-N..WILL BE CONTROLLED" DURING LONG.TERM COOLING OPERATION. TO% ENSURE THAT:THE-'PRESSURE-TEMIPERATURE.RESTRICTIONS'ARE.NOT VIOLATED."' ATTACHMENT -___ ..-.... ;..L -- ; I WI--D E -~ P-tP~ k(7 Y --- -rf .---I 9-1 .-1 I .1 ...I... ...... s.-' A V4, , 'I-V I 1. .. .. I-.^ .I.I I -.I ..II ,I I I-12>oo Ztj0o t 22co lu I V) / "&5b Zt oo~C..'fippvc. 81.7o ST=Y >5m7/u:!-,TESES NOCVVEip/~ 7.w-HP e-ala wIDVk / /..7___ I* p *-A..I~,rL: C- 0* ..* *fl~ I~ k .C .m v c ro4AZ*It-tA Ez r.k .r-..

3.7 CORE MELT. CONSIDERATIONS UNDER THE CONDITIONS OF NATURAL CIRCULATION, THE CORE WILL-BE SURROUNDED BY COLD- WATER NEAR 100 F. THE POSSIBILITY OF CORE MELT IS CONSIDERED,-

TO BE REMOTE UNDER TIHESE CONDITIONS. WITH THE CURRENT 'LOW DECAY HEAT'-RATE .AND WITH APPROPRIATE MONITORING' OF 'INCORE .THERMOCOUPLES, THERE WILL.BE SUFFICIENT EARLY WAqRNING SIGNALS TO -PREVENT CORE DAMAGE..A DETAILED DISCUSSIOX'OF CORE: .MELTING POINT, ASSESSMENT OF ORIGINAL FUEL DAMAGE CONDITIONS AND EARLY WARNING OSIGNALS IS .PROVIDED IN.ATTACHMENT 5.. BASED UPON 1THIS ASSESSEMENT AND THE USE OF DETAILED ACCEPTANCE CRITERIA: OR THE: TRANSITION TO- NATURAL CIRCULATION, IT IS'METROPOLITAN EDISON COMIPANY.'S VIEW THAT'NO 'PROBLMIS' EXIST WITH RESPECT TO THE POTENTIAL FOR CORE MELT.-:.I.I -,': .-.'b.-I r0.I f ..E i , 3.8 CORE THERMAL BEHAVIOR -THE CORE THERMAL BEHAVIOR FOR VARIOUS POTENTIAL MODES OF.OPERATION.IS DISCUSSED IN ATTACHMENT

4. ATTACHMENT 4 ALSO DISCUSSES THE USE'OF INCORE THERMOCOUPLES, CORE BLOCKAGE CONSIDERATIONS, THERMAL HYDRAULIC EVALUATION OF NATURAL:.CIRCULATION AND ANALYSIS OF.VARIOUS ALTERNATIVES.

BASED UPON THE.INFORMATION IN ATTACHMENT 4, METROPOLITAN EDISON COMPANY CONCLUDES THAT THE PLANT CAN BE SAFELY'OPERATED IN THE RECOMMENDED

LONG TERM-COOLING MODE..1. ..........: .,...-....,.1......i: , w.i s ..

-TE-2 CRITICALITY EVALUATION. Evlaios fcore subcriticaliiy ~and potentially critical fuel config ration were beguns sooafter the.TMI-2 incident. The.analsscvee rodsett of fuel configurations, rafiging fiom-the:1ftact core`> to :homo'geneous

solutioins dfL uranium- a a wter;'-:.Boroni~concentrations~necessarv.

to manti bctclty or the various postulated cozifigurations .were:Ide'ermined'. The. f ollowin'g' is -a~*.descripticon of ithe meth6d-s..of laialysis 'an d` iresu'lts 'frbm the criticality. evaltuations., 1.. Fuel -in..Core Region .. .:. ..-The analysis for the various possible 'configurations 'of ful-in. the coeezOn.was: diie notoaes (1) fuel,.rods:.intact and (2), successive t"slt pfrg of fu0l pellets I to an .iUltimate'. slab of pellets." 1.1 Fue 61Rods Intact'.~ .. .. .-.?PDQ-07 -calculations were performed for-the TNI-2 core -at the core'.burnupD on results are summarized in Tab lel.I .~ TABLE 1 Boron.Reguirements f or ThI Core' for Coli ~(70 0 F) :Shutdow'n.

.Temp~erature,.!F Control Rods, Keff. Boron, ppm..70 All Rods out -.95 2155.70, All- Rods :Out' .991795-70 .All Rods In ..95 1705:70 'All Rods In .99 .1385 Thebooncocetrtinslite~ao~e rebaedupon

.-the f oll'owing assu'mpticlis: a.; Guide':tubes, spacer grids and-cladding remai-n intact T efulith .basic structure as originally 'l6ade'd {nto the core -region.b. 1o credit aknor Lupdutrnable. Poison (LBP).c. 'Xenon full' -decayed. -~d. No credit takEin for- Samiarium ~buildup sinc'e dshutdown; equilibrium Samariu'At ho~t f ull'. power was as sutmed.--e. An. additional.. 1%. AK/K was included to6roid con servtive prediction at- 70 0 F.. 2 The values in Table .1 Are possibly n-conservative for higher. temperatures' beas h-core has a-positi~ve moderator coefficient of-48x10 KK 0 beas th: : o +8x10 AI/i at'2100 ppm boron, for. an intact 'core.~.Predicted keff 'value~s f or' the p~e~eni Itemperature and boron conditions based Qfl th e above as~sumpt~ions are shown in' Table 2.-TABLE. 2: -@0 Core Keff Values. f or Pres'ent. Conditions, .280 Fand 2100.ppm.Tem ...ture, F -Boron, .pp Contro:,Rods ., , -. 2.' ' 280. 210 All Rods Out ..O '280. 2100. All Rodsh .n92-* Predicted values reom Table, 1:using a +.8v f10r AK/K/F tmoderator -coef icient. at 2100 ppm boron. ..1.2 Fuel "Pellets: Slmpin.g* Criticality vastudies fowere ph rfme forful pelletsr fre fbrom cthecang dropping ~the pabove grids. Thegrids were successively. as.umeo .-r'p:' i ConeKf o. us~o E'eet.odins:8 F and 100ail resulting in :var~iou's slabs oCf ."sluMped 1' f uel: Atop .the:lower fuel.. s eget un'l ' .:..-a ', 'll-.the-f-'el. .ellets.r'ested.atoph:Q..: grid.KN-IV vrin2uliing 123.g XSDRN- cross-sec tion :sets were used. for All calculations...V. Tecr.w-odldin:'seven-symmetric planes -'ith -top and, bott'm ref lectbr" bu-nfntin xy.- -thus, a slab reactor: tlass calcul.hted for :160 F moderator-ttemperature, 'psi. :. (AnBetra pmconservatism. ewasf* .thetop .' ..plane..haiig,21 in-ches'.

. f'uel 1 as opposed' to.the -inch' actual)...

The initial axial geoety.... assumed.full .(21') pellet sin g a .s'ur1d 0 by borated water 7 ..This'-coniton e~eseitd. aUOj1 2 0voe' r o ' 307. Subsequent calculations ..ass:ed that's' thF f ueel sluSpped .i.'toa. ach.f the en.p' -ies, .-deined by the' spacer...grid, -a, U0 2/H 2 0 .-volu' eratio.~of 0 ,63 opccu-red wihi -the fu el- *a ter...-Th .-n, .o -.6...3 .mixture.- The assumed volume f rafctionof -6 is basead onpmeasured ..'packing, f raction data at- CNFP gil The calgulations .assumed 'BO'isotopics,`no control-: i, ii: ,-;n' -.utlut'"t n lag'sa'o'l ,tuth~ uel pellts rested onloy:b h vted rods'..no

.BPs, -n pi .... -.I e .Acalculatin was: also m.ade- assuming :that all s.ven planes'chad .lumoped ito one' lab.; Fuel pnrichme t for all casesnttas

'assumed:t:o- -.': .ithem0. aera .o the threeifuel batches, pie 2.60 w/o havin. ..Fo 'thosesonf f ions whichprodue a inc'ctal)'.emat the initial beorn , concentrationd f of -.210 p. ele (appro:dmatelevels in RC:systen)- wathe 'bo 'P ..., sue tht.a'h ful'lzae itoec:fthsvn. aes dernedwby increased r.ntgid a subcritical' array could be p0.edicted.' The -fuel arrangelmert .ous cases iusme show-m in Figure £1 ao the rbesUts are sudmacig d. ir : Table- 3. - TABLE 3 Criticality.:Calculations f or Four "Sluuipe~d"-Fuel Configuiations.'in Core Regtion Case.Keff. ~Case1 Case__ ..Case 3 Case .4 Boopm(Grids Intact) (2 Grids Fail) (5 Grids Fa'il) (To tal -Slump)2100 0.824 +/- .004 1.016 .005 1.062 .004: 1. 07 8+/- ..004 3000 .990 +/-.004 1.035 +/--.005 1.038 +/-.004 4000 -I- .992 +/-004 1.003 ..005 naditoal KNO :caIcu~laticn.w as made for the mnor credible -iutinof tiling.and the.fuel slumping Atop .fo, gd. F thssituation the ..fuel bel'ow the,-foiirth grid was assumed to besadng in its basic configuration. The a.rray was .'issumed

infinite in -y direction..

The slumped fuel .e -t n t ' eo was -.ased t' be packed :in -e -most optimum' fuel'/wateir-ratio b f :fue'l Vol ulme fraction' 0.55.Wherea, te 'standing fuelaipellets Fwere (ssumed.totbe-a t itial volume-*-fraction-of 0.307. Siinc e thi! KNO 'co de canno t' calculate~two. mixtures. of ~~fuel.with different c:ll. -h bove scena.rio was. .1:. wo .1.03 ases..:The first case assumed'fiormal celA lpitch (1c.l44 ) fbut: hd' to use ofarge : pellet 3.D. in the, slumped regionato proue the .55 volume f r'ction uel. -,-a -theit nai'ofgrtor.: Thcra 'sasmd niiei -ielto.Case 2 assumed-the omil.pellet..D n thie slumped region pwithacke ic tch of.1.12 cm'a a correspondily smaller pellet l.D.u'int thlowermintacutofuel: ), :of..2c and a. cresodinglysalrple ..'n'h'oe natfe-rods. Figuree2 shows thelfual configuration and Table 4 sunamrizes the 'results.TABLE 4 Criticality Calculations for.3 Grids Failing .q0 Case' Cell Pitch Boron, ppm K 1' 1.44cm 3000 280 1.000 2 .1.12 cm 3000 .280 1.019 .Comparing Case 2 above with'Ehe total slump Case 4 :from Table 1, the single, slab' is slightly lore reacti.ve. .An assessment of the inherent conservatism and.nonconservatism of. the KENO studies (Tables 3 & 4) is presented below: 1. No fuel depletion considered. _Fuel depletion will reduce K.by.-. -2.5%. AK/K.2. No radial leakage considered. Radial leakage will reduce Ka by " -3.0% -K/K.3. No credit'. taken for Ag-In-Cd:Control Rods. '4. The'total.'slump case (slab) assumed'a Volume Fraction of fuel equal to 0.63, the packing fraction. llowever, this is not necessarily the optimum configuratio n; for criticality (Section'. ). If thetmost optimum. fuel/water'ratio. is fonned, the configuration may be more reactive'by " +2.5% AK/K. 4.1.3 Fuel Pellets: Sphee 1.3i iue Peltf-pee: .' 'r ',~ ssmt 'n .t.a Calculations were performed to predict'the criticality for the assumption that: -: all the fuelpellets collapse' to. form an optimum-spherical configuration in the'reactor vessel plenum The multiplication'factor for an infinite array of non-depleted fuel-pellets in the optimum water-ratio for the core average enrichment 2.6:w/o U-235 is presented 'in Table .5 Also shown in Table 5 is the- expected leakage.reactivity 'for thehypothetical sphere.0 Enrichment Temp, F Boron, ppm Optimum.VF Fuel KX P Leakage w /o _U 235 ' '._-.- _.____________' 2.6 280.. '2100 0.52 1.91':-1.4% AK/K'2.6 280 3000 ' '0.63 1.028-.-1.6% AK/Kj: These data.demonstrate that after subtracting the.fuel depletion reactivity -(-2.5% AK/K) and'the expected leakage, a hypothetical sphere'can be-critica1 at 2100 ppm but'will be subcritical -at a-boron -concentration of 3000 ppm..2. Parameter Study -Infin'ite Media .Heterogeneous mixtures of fuel 'andwater in an.infinite array- were analyzed': with the NUL-IF code. The volume fr'action of.'.fuel, 1fuel .particle size,-boron concentration, and temperature were .varied. The calculations -were performed** for a fuel. enrichmen. of 2.6 '/oU-235,..co'rresponding to'.the average'fuel. enrichment in the core. No credit:..was taken for fuel buirhup:-or fission product'.buildup. -.Figur 3 shows reactivity as a function of the'uranium volume fraction for.several boron concentrations ind 2'different:-temperatures'. The: optimum fuel/ -water.ratio.-increases'.with increasing boron.concentration. The.-most likely ' .-: fuel volume 'fraction for inta'ct fuel pelle't-s settling in' a system cavity has'.:been experimentally. determined to be:.. 63. Figure 4 shows. reactivity as. a., function of .fuei particle size for a fuelyo1ume fraction':of ..63.. .As- can be: seen.by comparing Figure's 3 and A4., reactivity is-much'more sensitive'to 'the':. volume fraction of -uranium.-in the system -than to .particle size. For fuel-volume fractions in 'the range where K pcanbe'greater:than'1.0 with 2100 pp boron in the system, :intact -fuel :pellets uniformly."distributed-in the system were found to. be i2ore reactive th an .equal amount of-'fuel disaersed: in smaller. sized particles --oi horaogeneously .mixed-.with ithe coolant. --Figure 5.shows reactivity'as a function. of-boron concentration for. nium.volume fractions:of .;52 and'.63 at.213 0 F At lower boron~cncentrations, the lower '.uranium volume fractions. a re -zare reactive at' higher boron concentratins,. the larger uranium volume fra' ti6o n are mo re reactive'.- Figure 6::shows'l-reactivity as '.function: of' boron concentration .for a .-uranium volumefraction. of .63 and 3 different'particle sizes. -All the above'calculations were~performed at 1000 psia'. A drop in system -pressure to 300 psia.wouldbe -equivalent.'to a 10 0 F.:rise in .the'moderatr '. 'temperature. 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..../. ..... ... ............ /7_a, ..,, ,1_ ...:-Babcock&Wilcox April 10, 1979 wI A SUMIM4ARY OF NATURAL CIRCULATION ALTERNATIVES FOR LONG-TERN CORE COOLING AT TMI-2 .II prepared by-. -B. A. Karrasch. TABLE OF CONTENTS* INTRODUCTION

SUMMARY

DISCUSSION STEADY STATE RESULTS TRANSIENT ANALYSES -OTSG TEST PROGRAM ACCEPTANCE CRITERIA DURING OPERATION' 'RECOMNENDATION APPENDICES REQUIRED INSTRUMENTATION REFERENCES i 1 INTRODUCTION: The TMI-2 long-term cooling mode proposed by B&W utilizes RCS natural circ-ulation core cooling with the steam generator secondary side in a solid*flowing water condition for heat removal. This. ultimate decay heat removal mode is a key feature of the following proposed sequence of events to achieve a stable, cold safe shutdown condition at THI-2.Phase I: Reduce RCS temperature to approximately 230 F.by steaming the IA OTSG through the turbine bypass system.with one RC pump running and RC pressure controlled. to. a value greater than the pump NPSH using the pressurizer in a normal. mode.Phase II:*A With the A OTSG steaming, the X OTSG (and closed secondary system yet to be installed) will be slowly filled-solid with water and a transition will be made to remove RCS A8 decay heat with the > OTSG solid. The KOTSG will be isolated and the RCS temperature will be reduced to approximately 100 F with the if OTSG. Reactor coolant flow and pressure conditions willremain the same as Phase I.Phase III: The A OTSG (and closed secondary system to be installed) will be filled solid with water and a transition made to remove RCS decay heat.with both A and B steam* .generators flowing solid with 1000F feedwater. Reactor coolant flow and pressure conditions will remain the same as.Phase I..I V --..w .--1-.q Alternate to Phases II and III: Phase IV: The B OTSG will remain isolated and the A OTSG will continue to remove RCS decay heat *during the transition from'a steaming secondary at 230 F to a solid-water secondary at 100 F. This scheme would not utilize the B OTSG; however, the transition operation is more difficult with respect to steam line water hammer and maintenanze of a stable RCS temperature and pressure.With the reactor coolant system at approximately 100 F using normal RC pressure control and secondary side heat removal with a solid system (between 3000 and 5000 gpm 0 at 100 F)' the reactor coolant pump will be tripped and natural circulation core cooling will commence.Acceptance criteria for core cooling will be established and long-term cooling of the core will be maintained with natural circulation.

Phase V: With natural circulation for core cooling and a solid secondarysystem for OTSG heat removal, RC pressure can be reduced to a minimum value required to maintain the RCS in a sub-cooled condition.

Our plan is to fill the primary system solid, including the pressurizer, and maintain pressure control with a makeup pump designed for such an application. To maintain a stable sub-cooled margin, we envision a long-term RCS pressure between 20.and 50 psia. 'During the past week, analysis and testing has been underway at NPGD and..the Alliance Research .Center-to-define and understand ..the various'alternatives available for core heat removal with natural circulation. The analyses were-2-7 --directed toward obtaining data to define an optimum long-term cooling mode.Several of the important considerations include: 1. Core natural circulation cooling, for various core AP configurations, with one or two OTSG's in service.2. OTSG natural circulation cooling performance with various secondary side water flowrates to the unit through the main or auxiliary feedwater nozzles.3. Expected transient performance during the transition from forced to circulation including specific acceptance criteria for the operator determine if adequate core cooling is achieved.natural to-3-...

SUMMARY

The preferred mode for natural circulation core cooling is to use both OTSG's solid on the secondary side, with a flowrate of 3000 gpm, entering ...the OTSG through the main feedwater nozzles and exiting the unit through the steam outlet nozzles. -This mode will provide a maximum. core flowrate (> 800,000 lb/hr), a min~imum core AT.(< 30 0 F), and a' minimum reactor coolant 0 e average temperature (< 1200F for a 1000F OTSG feedwater temperature)., The secondary. side OTSG cooling is a stable, forced convection mode, which tratnsf ers all the primary system Energy above a tube elevation of 30 feet, thereby providing..a high column of -cold-water for enhancing.'.the primary side.natural circulation. This mode provides a driving head similar to that obtained with the OTSG steaming with a secondary side level at 30 feet.The solid secondary side mode of operation has a' distinct advantage over a steaming mode in that a much lower reactor' coolant system temperature can be achieved. The solid configuration will result in an RCS temperature very close to the OTSG f eedwater temperature (approximately. 100 F); the steaming mode of operation can only obtain RCS conditions equivalent to the saturation 0 temperature at the lowest achievable' steam pressure (approximately 230 F).In addition, the use of-the main nozzles for .OTSG. feedwater. addition has been shown to yield a predictable and uniform 'primary system heat 'removal suitable for natural 'circulation;. The use-of the'auxiliary nozzles *for OTSG feedwater addition,' with water exiting the main nozzles, should. also'remove the primary heat at an elevated point in the unit. However, the flow distribution and uniformity of cooling is uncertain~and the feedwater f lowqrates are limited' by' systrei design and OTSG tube crossf low velocity.' concerns. In addition,' major, secondary plant modification would be required to implement reverse flow throigh the'.OTSG miain feedwa er nozzes.Tetg -'-4 performed on the 19-tube steam generator at.the Alliance Research Center confirms that feedwater addition through the main nozzles with water exiting the steam.outlet nozzles is the preferable mode for natural circulation. The advantage of using both steam generators instead of only one is an increase in the core natural circulation core flowrate of 10 to 20 percent 0-and a decrease in the core outlet temperature of about 5 F. Extensive analysis has been independently performed at NPGD to confirm that the difference between using one or two OTSG's is not significant from a natural circulation' standpoint; one OTSG in service will provide adequate core cooling. We believe, however, that t:he uncertainty in local core conditions and cooling requirements, .the need for heat exchanger redundancy in the long-termcooling mode, and the ease of transition and operation with a solid water secondary for two OTSG's versus one, makes operation with two loops a superior mode.The effect of a greater core resistance on the natural circulation cooling capability has been evaluated-and deemed -acceptable. A .core.resistance of 60-times the normal value has been assumed in the calculations, and the* reported results are acceptable for either one or:two steam.generators in operation. The difference between a normal core resistance' and a -core .resistance 60 times normal (indicating a significant blockage) is a factor of two in core flow. and AT. This favorable result is due to the offsetting effects of system resistance, flowrate, and temperature difference to sustain a stable natural circulation'condition..- Expected transient performance -during the transition from forced primary system flow to-natural circulation, is predictable and. stable. From an initial condition with the RC pump'running and primary and secondary temperature: approximately 100 F, a stable natural circulation, condition will -be achieved.within a half hour following the pump trip. The'cold.leg temperature will: decrease slightly (due to the pump power loss) and remain stable at. about - 0 0 100 F. The core outlet temperature will increase by about 20-30 F within 10 minutes and be observed on the-hot leg RTD in less than 20 minutes.During the first hour after the pump trip, the reactor vessel heatup with no primary system flow would only be 100 0 F. Acceptance criteria during the first hour of natural circulation will be provided to the operator and primary system pressure will be maintained to assure that the reactor core outlet 0 temperature remrains 100 F sub-cooled at all times. When the operator observes the increase in hot leg temperature indication; a stable natural circulation condition will be confirmed.

1 DISCUSSION A. Steady State Analysis The results of the steady state natural circulation analyses performed to date are presented in Table 1. Four different reactor configurations were evaluated to determine the sensitivity of various conditions and assumptions on the natural circulation core flowrate and core temper-ature drop. The configurations studied include: 1. Two loop operation with both steam generators steaming at 230 F (20 psia) at a 30-foot secondary side level (95% on operate range).'This configuration is similar to that' which has been tested on the Oconee Units and forms the basis for a considerable amount of analysis at NPGD.. These cases have been used to provide a bench-mark on the OTSG heat transfer characteristics for development of a driving head and for confirming RCS loop AP characteristics.

Figure 1 illustrates the sensitivity of the core natural circulation flowrate with loop AT (the driving head gain) and loop pressure drop (the driving head loss). The flowrate will seek a stable natural circulation condition based upon the loop AP and the resultant core AT. The key to obtaining a maximum flowrate is to remove the primary system heat (i.e., change th to T. ldat as high an elevation as possible in the steam generator. Our testing and analysis confirms that the primary heat is all trans-ferred above the'liquid/steam interface (i.e., the level) on the secondary side of the OTSG. The calculational results presented conservatively assume that the primary system temperature change occurs as a step change at the height of the OTSG operate range level.a

The effect of increased..core AP has also been evaluated to determine the core flowrate and temperature drop sensitivity.

The following types of analyses have been performed at NPGD to conclude, that the TMI-2 core resistance in its current configuration could be as high as 60 times the nominal value, indicating a high degree of core blockage: a. Core AP calculations based upon a postulated core configuration.

b. A comparison of RCS flow meter readings, with one.pump running, before and after the TMI-2 incident.c. A conservative estimate of core flowrate and pressure drop in the current TMI-2 core configuration.using the actual decay heat level and the difference between the cold leg'.temperature' and the core outlet temperature as determined by the core outlet thermocouples

[i.e., core flow decay and AP (core flow) ].AT iThese analyses have provided a range of core AP values which have been included in the evaluations described in Table 1.The effect of increased core AP on the natural circulation

flowrate is illustrated on Figure 2. The analyses have shown that the natural circulation flowrates are adequate with the core'in its current configuration.

2. Single loop operation with OTSG A steaming at 230 F (20 psia) at* a 30-foot secondary side level-OTSG B isolated.

This. configuration has been evaluated to provide a comparison of:two loop versus single loop operation. The single loop calculations confirm that* the net core flow will be 10 to 20% less in this configuration than with both steam generators in service.. The-resultant core AT will. increase about 5 F (depending upon the decay heat level) and is still acceptable for core cooling. These analyses were performed to confirm an acceptable condition should an emergency situation require an immediate transition to natural circulation prior to the plannrd sequence to a solid stean generator secondary side.3. Single loop operation with OTSG*A inma solid secondary.side mode with water addition through the main feedwater nozzles. Steam generator heat transfer analyses and testing have confirmed that a.3000 gpm feedwater flowrate to the main feedwater nozzles will provide a primary to secondary heat transfer characteristic similar to that achieved with the OTSG steaming with a 30-foot .water level..... The majority of the heat removal occurs above the..30-foot level. in the OTSG with a 3000 gpm floirrate. If_. the. flowrate is increased to 5000 gpm, the driving head for natural circulation is further improved to about 35 feet. The calculational results confirm that adequate natural circulation flow and core AT can be obtained with a single steam generator operating inma solid condition. Additional analyses were performed in this configuration to'determine the effect of reduced dore decay-heat levels. These cases were run at 2 and 3 MW to provide a'comparisbn with values of core flow and AT at 5 MW. As can be seen from Table 1, .natural:circulation core flowrate and core AT are both reduced for-.lower decay heat values, and core cooling remains acceptable.

4. Two loop operation with both steam generators in a solid secondary mode with feedwater addition through main nozzles at 3000 gpm. .This is the preferred mode for long-term cooling at TMI-2 and the results are very similar to the two OTSG's steaming case.-Again, the solid flowing water secondary system at 3000 gpm induces a high heat transfer interface in the OTSG's and acceptable core natural circulation' cooling is achieved.'The steady state natural circulation analysis has resulted in the following conclusions with regard to long-term cooling at TMI-2: 1. Adequate core cooling with natural circulation can be achieved with either one or two steam generators in service.I 2. An increased core resistance due to blockage decreases the natural circulation flowrate and increases the core AT. However, it has been shown that acceptable core flow and.:AT can be maintained with significant increases in the core resistance due to blockage..

3. 'Adequate natural circulation flowrate can be achieved with the steam generator(s) in a steaming or solid mode if the effective heat transfer height is maintained at 30 feet or greater using a high level for steaming (30 feet) or a high flowrate for solid (3000 gpm).4. Adequate natural circulation cooling can be. maintained at reduced core decay heat levels.-lo .i -- 0 B. Transient.Analysis.'

The results of the forced flow transition to natural circulation cooling are presented in Figures 3 and 4. Transient analyses were performed at core decay heat levels of 2 and 3 megawatts to better.. ,* understand the time dependent: responses of core flowra'te.and'temper- ,-ature change following the loss of forced cooling. The bases for the analyses are as follows:* Core Power -.2 and.3 Megawatts Reactor Coolant Pump Trip atTime 0 OTSG A Solid with 100F Feedwater into the Main Nozzles at 3000 gpm OTSG B Isolated Core Resistance Factor -60 The transient responses of core flow and temperature confirm that a smooth transition .to natural circulation is achievable. Following.the loss of forced flow, the reactor vessel heatup slowly induces a temperature gradient between the reactor vessel and upper OTSG and natural circulation occurs with no operator action. The core flowrate reaches a minimum about 1 minute.into the transient and reaches a stable condition between 10 and 20 minutes. The core outlet temperature begins to increase and reaches a maximum value 4 to 5 minutes into the transient and a stable condition at about 10 minutes. There is about a four minute time delay in the response of the hot leg temper-3 ature measurement due to the approximately 1000 feet in the reactor vessel upper plenum and hot leg piping. The cold leg temperature drops slowly to closely match the OTSG feedwater temperature due to the loss of the approximately 5 megawatts of pumping power. .The transient natural circulation analyses have resulted in the following conclusions with regard to long-term cooling at TMI-2:-11-: __ _- -__ ___ _. ___ ---. .- ____ .- __ .__ .- -__ -.- ------_ __ -__1. A smooth transition from forced flow cooling to natural circulation can be achieved by tripping the reactor coolant pump and observing core outlet temperature. There is no reason to slowly reduce the RC pump speed for a more gradual transition to natural circulation.

2. The reactor coolant system flow and temperature will reach an equilibrium value within the first 1/2 hour of the transient; the'response of the hot leg temperature measurement-occurs within 5 minutes after the core outlet temperature changes.

C. OTSG.Test Program. The ability to achieve and maintain a stable natural circulation flowrate is.dependent upon the elevation difference between the heated core outlet temperature and the transition to cold leg.temperature in the OTSG tubes.. This transition point in the OTSG.is in turn dependent -.upon the heat transfer-characteristics of the unit'. If the primary to secondary heat transfer can be obtained at-a-high elevation in the -OTSG, the driving head from the density'difference will be improved and natural circulation.flowrate will increase.The primary to secondary OTSG heat transfer mechanism,while in a steaming mode is boiling at or about the level of-the secondary side water.Extensive analysis and testing of the OTSG in the "pot boiling" mode has confirmed that the primary system temperature transition-occurs above the level of the "boiling'pot." All calculations performed with the OTSG in a steaming mode conservatively assume that the primary system cold leg temperature is available for driving the natural circulation flow at the 30 foot level. .Use of the steam generator as a water to water counter flow heat.-; exchanger is a more desirable condition to obtain during a long-term decay heat cooling mode. The primary temperatures can be maintained

much nearer the temperature of the incoming feedwater to the OTSG.In order to determine OTSG characteristics-in a solid mode, a test -program was conducted at the Alliance Research Center on a 19-tube, full-length, steam generator.

A natural circulation flowrate of. 700,000 lb/hr was simulated on .the primary side and forced secondary side cooling was injected into the main feedwater nozzles; flow-exited the unit through the steam outlet nozzles. Feedwater flowrates were varied from a scaled value of 100 gpm up to 5000 'gpm. - The results of this test program are presented on Figure 5, a plot.of feedwater flowrate versus the OTSG heat transfer elevation. Heat transfer elevation is defined as that level above which all primary system heat is transferred to the secondary system fluid. That is, the height at which one can assume the primary cold leg temperature is available for driving natural. circulation. The figure shows that a heat-transfer elevation of 30 feet can be obtained-if the feedwater flowrate is at 3000 gpm or higher. A 30 foot elevation head in the OTSG primary is adequate to achieve natural circulation as demonstrated by the calculations in the previous section...iI..-i4-D. Acceptance Criteria During-Operation.The success or failure of natural circulation as a core cooling mode depends upon the value of the core AT that can be maintained. The key objective during plant operation in this mode is to.maintain a primary cold leg temperature as low as possible and observe the resultant hot leg temperature; The acceptance criteria for success of the natural -* : circulation mode is to maintain the hot.leg temperature below the sat-uration temperature which would -cause bulk boiling. --Figure 6 illustrates -*the proposed NPGD criteria for natural circulation: to maintain a 1000F sub-cooled margin to bulk boiling using the plant instrumentation ' in its current degraded state. The large errors which have been -imposed* on the pressure and temperature instrumentation make it imperative to.*keep the RC pressure as high as possible at the time of pump trip. This-will allow a large hot leg temperature increase to occur before boiling -and assure a reasonable time period to achieve a stable natural circulation. If the RC pressure is maintained at 500 psia, the hot leg temperature --0 0' can reach 340 F (from its initial condition of 110 F) before action -must be taken.An analysis of the 'reactor vessel was performed to determine the potential.- heatup rate with zero flow into the vessel. This analysis provides a bounding bulk fluid heatup rate to indicate the amount of time available to the operator to take action before a boiling condition could occur.The high probability of achieving a stable.natural circulation condition indicates that such a reactor vessel hetup could never occur. The analysis is provided to show that the operator has-at least one hour to confirm natural circulation before any action must be taken. RECOMIENDATION:

Natural circulation has been shown to be an acceptable m~ans of-heat removal for long-term cooling at TMI-2 with the core in its current config-* uration. Use of either one or two steam generators is feasible if the prpe:.scondary side heaf transfer charactrsisate established and maintained to remove the primary energy near the top of the OTSG. 'In addition, the expected transition process from forced cooling to natural circulation will provide *a continuous and stable core cooling cond ition which can be monitored and controlled by the plant operator.B&W, therefore recommends-that a planned transition to natural circulation core cooling be implemented at TMI1-2 as soon as the degassing process is*** completed.

Both steam generators should be utilized in a solid flowing water condition with approximately 100 F feedwater at 3000 to 5000 gpm entering through the main feedwater nozzles. The-sequence of events for..this transition, as described in the Introduction of this report should be as follows: 1. Reduce RCS temperature to 2300F with a OTG steaming.2. Slowly fill OTSG B solid with water and begin removing primary system energy with the B OTSG by gradually increasing feedwater flow-until a stable condition is reached at 200-230 F. When a stable conidition has been established, isolate OTSG A.3. Reduce'the.RCS temperature to approximately 100 F by increasing-the feedwater flowrate to OTSG B. 'Fill OTSG A solid with water and prepare for operation. +16-

4. Slowly begin. feeding OTSG A with 100 F feedwater and establish a stable condition with both steam generators removing decay heat with a 3000 gpm feedwater flowrate.5. Establish'R.CS natural circulation as follows: a. Throttle feedwater flow to both steam generators-to establish approximately 25 F AT between feedwater temperature and OTSG secondary outlet-temperature..

b. When a stable condition has 'been'establsihed, trip the running reactor coolant pump and increase feedwater flowrate to both OTSG's to 5000 gpm within 3 minutes. Maintain at 5000 gpm.c. Maintain RC pressure at the initial condition value and observe both A'and B hot leg temperatures.

d. Compare the-hot leg temperatures to the acceptance criteria on Figure 6. If the temperature exceeds the limiting value, start a reactor coolant pump.e. When stable natural circulation conditions have been achieved, reduce'RC pressure to the proposed long-term cooling value between 20 and 50 psia..The above sequence of events will establish a stable and safe natural* circulation condition for long-term cooling at TMI-2. All starting or stopping of reactor coolant pumps should be avoided until the pump is tripped to induce natural circulation.

In addition, B&W recommends that the sequence of'ev'ents be implemented in a planned and.controlled..manner, i.e.,..we should not wait for a complete-failure of allifour RCP'-s before establishing. natural circulation. We should, however, have an alternate decay-heat removal system installed and ready for operation prior to the transition to natural circulation. ' APPENDIX A INSTRUMENTATION REQUIRED TO ACHIEVE AND MAINTAIN LONG TERM COOLING.I r ........; I ..1 .....INSTRUMENTATION REQUIRED TO ACHIEVE I. REQUIRED TO CONFIRM Primary System ITEM I 2 3-4 5 II. REQUIRED TO MONITOR A. Primary System ITEM*1'2;3 4.THE INITIATION OF NATURAL CIRCULATION MEASUREMENT Reactor Core Outlet Temperature (Incore TCWs)Loop A Reactor Hot Leg Temperature Loop B Reactor Hot Leg Temperature .Loop Al or A2 Cold Leg Temperature Loop Bl or B2 Cold Leg Temperature RANGE OF INTEREST O-700F 0-550F 0-550F 0-350F 0-350F AND MAINTAIN LONG TERM COOLING DESIRED ACCURACY+ lOF+ 1OF+ lOF+ lOF PAGE l 3 COMMENTS BACK-UP MEASUREMENT Item I.2; I.3 rtem 1.3 Item 1.2 Item r.5 Item 1.4 LONG TERM NATURAL CIRCULATION (IN ADDITION TO ITEMS IN I. ABOVE)MEASUREMENT Pressurizer Level Pressurizer Temperature ..A Loop A Reactor Coolant Pressure Loop B Reactor.Coolant Pressure RANGE OF INTEREST 0-400" 0-500F 0-1000 pslg : 0-1000 pslg DESIRED ACCURACY+ 40"+ lOF+ 50 psi+ 50 psI BACK-UP MEASUREMENT Item II.A.3 Item MM.A.4 Item Ir.A.3 COMMENTS Comp. I.D. 0387, 0388 (Not required for solid primary.)Comp. I.D. 0389 (Not required for solid primary.)Comp. I.D. 0398, 0399 Comp. I.D. 0400 PAGE 2 of .INSTRUMENTATION REQUIRED TO ACHIEVE AND MAINTAIN LUNG TERM COOLING B. Secondary System ITEM MEASUREMENT. RANGE OF INTEREST 0-600" 1.32 4 6 7-8 10-11*12-13 14.Steam Generator A Level Steam Generator B Level-Steam Generator A Outlet Pressure Steam Generator B Outlet Pressure Steam Generator A Main Feedwater Flow Steam Generator B Main Feedwater Flow Steam Generator A Start-Up Feedwater Flow Steam Generator B Start-Up Feedwater Flow Steam Generator A Outlet Temperature Steam Generator B Outlet Temperature Steam Generator A Feed Temperature Steam Generator B Feed Temperature Steam Generator A downcomer Temperature Steam Generator B Downcomer Temperature DESIRED ACCURACY+ 30"+ 30" 0-600" BACK-UP MEASUREMENT 0-300 psig 0-300 psig 0-7000 gpm 0-7000 gpm 0-500 gpm 0-500 gpm 0-250 F 0-250 F 0-150 F 0-150 F 0-150 F 0-150 F+/- 15 psi+ 15 psi+/- 50 gpm+ 50 gpm+ 10 gpm Item T.4.Item 1.5 Item II.BD1 Item IIB.2 COMMENTS Comp. I.D. 0009 (Required only if steaming.) COmp. 1.D. 0001 (Required only if steaming.)(Required only if steaming-.)(Required only if steaming.)(Required'only for solid secondary.)(Required only for.solid secondary.) Comp. I.D. 0491 Comp. I.D. 0492 Comp. T.D. 0469 Comp. 1.D. 0470+ 10 gpm+ 2 F+ 2 F+ 2 F+2 F+ 2 F -+ 2 F+2F.Item II.B.10 Item II.B.9 INSTRUCTION REQUIREMENTS TO ACHIEVEA& MAINTAIN LONG TERM COOLING.III. Additional Required Measurements ITEM MEASUREMENT 1 Pressurizer Heater Status (Banks 1,2,3,4,5) 2 Electromatic Relief Block Valve Position Status 3 Pressurizer Vent (EMO to Quench Tank) Valve Position 4 Reactor Vessel Boron Con-centration 5 Makeup Flow 6 Makeup Tank Level.7 Makeup Boron Con-centration 8Letdown Flow 9 .. Letdown Temperature 10 Borated Water Storage Tank Level 11 -Borated Water Storage Tank Temperature 12 Borated Water Storage Tank Concentration 13 Steam Generator A Activity 14 .Steam Generator B Activity.15 Heat Sink Temperature .16 Heat Sink Pressure RANGE OF INTEREST 0-540KW Per Bank Open; Closed Open; Closed Status.0-500Oppm 0-200gpm 0-100" 0-12000ppm 0-1OOgpm 0-200F 0-100'0-200F 0-5000ppm DESIRED ACCURACY+ lOOppm+ 5gpm+ 10"+200ppm+/- lOgpm+ 20F BACKUP MEASUREMENTS Page 3 of 3 COMMENTS Not required for solid primary Sample None None Sample I.I iI I i.I I I I I I I i I I.i.I I.i .Item III.9 None Redundant Sensor Comp. I.D. 0346+ 20F Level Level 0-200F 0-5psi 0-300psi*+ lO~ppm+ lOF:+ .0.5psi+7 2Opsi Sample Sample Sample 17 Heat Sink Level (If Applicable) I I I.I I , APPENDIX B .:.. ...., .... .. ..... ..... ...............,..LIST OF REFERENCES _ __, .:......ThiE: Appendix will be included in the final report and .: include a complete list of all calculations and related ;test: data and backup material for the information contained in this report. ..: ......,. : .. ': --....* * :.. .-.......... .... ..-.... ..* .. ...* *-.......... R........................ ....... ............-.. , .:* .. .....:-...' ' -.' .' .*- ' : '-......-.--l l ........-.-' .*.: T'le 1 SUMIARY OF NATURAL CIRCULATION ALTERNATIVES Reactor. Configuration Core Flow Core AT T Thot 20,i , 6 , ..Two OTSG's Steaming at 230 F With 0.8xlO 1.2xlO lb/hr 15-25 F 230 F 245-255 F a 30'.Secondary Level 60 Times Normal Core Resistance 0.8 -6.2x1. lb/hr 15-25 F 2300F 245-2550F 10 Times Normal Core Resistance .1.1 -1.6x10 6 lb/hr 11-19PF 2300F 241-249 0F Normal Core Resistance.- 1.5 -2.3xlO lb/hr 8-13OF 230 F .238-243 F 0 One.OTSG Steaming at 230 F With .6 o 9 30' Secondary Level -.0.7 -l.lxlO lb/hr 20-30 F 230 F 250-260 F (60 Times Normal Core Resistance). Two OTSG's Solid With 100 F 6 \ .Feedwater at 3000 gpm 0.8 -1.2xlO lb/hr A 15-25 F 105 F 120-125 F (60.Times Normal Core Resistance) .0. ...6 ..One OTSG Solid With lOO F .0.7 lilxlO lb/hr 20-30 F 105 F 125-135 F Feedwater at 3000 gpm (60 Times Normal Core Resistance)

3-6 0 0 11-20.3MW Decay Heat 0.7 -l.lx10 6* lb/hr 10-20 F 103°F 113-123 0F 2 MW Decay Heat -0.6 -l.OxlO lb/hr. 8-15 F 1010F 109-116.F.1 I.:.x)L tv t..f. ed)e9,M: -. ....3 11;14l7-I) Lr II I I

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3. 3. 1. DESCRIPTION.

OF CADDS.NATLURAL CIRCULATION MODEL THE CADDS DIGITAL COM4PUTER CODE IS DESIGNED.TO ANALYZE REACTOR TRANSIENTS,-WITH OR WITOTSRM NAHTEROGEN US PRtSSURIZED WATER:REACTOR.- IT.:.SOLVES THE TIME DEPENDENT-NEUTRON KINETICS EQUATIONS IN CONJUNCTION WI TH A 'THERMAL HYDRAULIC SOLUTION FOR AN AVERAGE FUEL IIN DURING.A REACTIVITY.-TRANSIENT.. THE S IMULATION INCLUDES THE MAJOR FEEDBACK MECHANISMS ASWELL AS.DETAIL'ED SINGLE--PRASE NUCLEATE BOILING, TRAINSITION, JAND STABLE FILM'BOILINTG CORE HEAT TRANSFER MODELS. -THE ENTIRE REACTOR COOLANT LOOP, INUCLUDING THE:PRESSURIZER, .IS SIMULATED... THE STEAMGENERATOR M.ODEL.IS -INCLUDED' TO' EVALUATE THE EFFECT OF FEEDWATER'VARIATIONS ON THE STEAM GENERATOR As WELL.AS-THE PRIM.&RY-SYSTEMA RESPONSE..-THE CADDS COM4PUTER CODE DESCRIBED ABOVE NORMALLY REQUIRES THA&T THE REACTOR'COOLANT-LOOP FLOW HISTORY BE SPECIF].ED AS PART OF THE CODE :INPUT.- HOWEVER, A RECENTLY-COMPLETED MODIFICATION TC) CADD ALLOWIS-FLOW COASTD0WN AND NATURAL CIRCULATION TO BE-CALCULATED WITHIN THE. PROGRAX. THE DEVELOPMENT OF THIS..-MODIFICATION'FOLLOWS_ ______--A SCHEMATIC DIAGRAM-OF THE PRIMARY LOOP IS SHOWN IN FIGURE 1.'. CONSIDERING'THE-FLOW AS ONE-DIMENSIONAL,: THE. MOMENTUIM -EQUATION. IN THE AXIAL DIRECTION x'IS..2 3P DF<.(P.) (P + .Fg 1 ata/x a t --..'WHERE F'REPRESENTS FRICTION/FORM, LOSS 'FORCES AND Fg IS THE GRAVITY FORCE.' INTEGRATING EQUATION (1) ALONG THE.PRIMARY SYSTEM IN THE DIRECTION OF FLOW.FROM THE PUMP DISCHARGE TO THE PUI2 SUCTION YIELDS 6. 6 2 66~a 2)_dx(2)0 ;0 .................................... 00l Pdx ' aa(JP41) K Fl ,(dx + l J2 d + Fdx. +..W -.at -.':,, .O, \ Z , SUBSTITUTING FOR F AND F AND EQUATING THE PRESSURE INTEGRAL TO AP OBTAINS-f 6; J -6a (P u)d. + a(dx + Ix -PuM a x g 0 a..1 t Pg*dX f ...eV (k +f L *.. .P ..uju WHERE cv.IS THE TOTAL NUMBER OF CONTROL VOLUMES. CONSIDERING:THE FLOW.RATE TO BE SPATIALLY INDEPENDENT, EXPRESSING VELOCITY (u) IN TERMS OF SYSTEI FLOW. RATE (W) AREAS (Ai) AND.DENSITIES( AND SUMMING OVER ALL CONTROL VOLUMES YIELDS ,.. ,p,, 2 r26 , wc~v Li c~v.: Pump .-. 1 +. .-..0 .L ...0 .+K'-+ p(4)I ~i ~29 2 ..*~~ .-: ..:, USING A SEMI-DIPLICIT SOLUTION TECHNIQUE, EQUATION 4 CAN BE WRITTEN AS...;. +l wncX Li =wI~ .__ flfl l t iAigo pump go PA 2*Vwe -e -L , cv -A JkL n Ll cv 1L.0 0 2 ,f (REARRANGING AND NEGLECTING THE ACCELERATION TERM PRODUCES.v L'.' Cv ..WI .1 *1 ,i i +. &PP* .* cv L j jcv .Li* -.i + 2g 12 (Ki+ fiD ) i io~ oi0A-a , -V .L..THEREFORE, THE SYSTEM FLOW RATE.CAN BEDETERMINED BY EQUATION 6 AT NY .TIME STEP-DURING THE TRANSIENT. THE MOMENTMI'EQUATION IS-COUPLED TO THE* .: CADDS.ENERGY EQUATION THROUGH THE FLUID PROPERTIES WHICH,'ARE'FUNCTIONS OF THE SYSTEM PRESSURE AND THE ENTHAI.PY. 'THE CALCULATION OF FLOW'RATE IS'CON-,* TAINED IN THE CADDS SUBROUTINE, NATURAL. '* 3.3.1,.1. COMPARISON OF-NATURAL CIRCULATION MODEL PREDICTIONS TO ANALYTICAL ... RESULTS: TWO SPECIAL CASES WERE USED TO VERIFY THE CALCULATIONS OF THE NATURAL SUB-.ROUTINE. EACH CASE MADE USE OF.ENOUGH SIMPLIFYING ASSUMPTIONS TO ALLOW THE' -) HAND CALCULATION OF FLOW (OR FLOW VERSUS TIME) WHICH COULD THEN BE COMPARED TO NATURAL PREDICTIONS., -^ THE FIRST CASE SIMULATED THE FLOW COASTDOWN OF: AN NSS PRIMARY LOOP. IT WAS ASSUMED THAT THE PUMP SEIZED AT THE BEGINNING OF THE TRANSIENT (AP. = 0) AND TEAT THE GRAVITY TERN COUD BE NEGLECTED SINCE INERTIA TEPEMS pUMp DOMINATE IN THE INITIAL STAGES OF A'.COASTDOWN TRANSIENT. THE NATURAL.SUBROUTINE RESULTS ARE COMPARED TO ANALYTICAL VALUES IN TABLE1..11 -FI S , ..THE SECOND CASE USED THE SAME PRIMARY LOOP AS 'SE FIRST CASE-AND ASSUMED:* .A LINEAR'DISTRIBUTION OF DENSITY' INTHE CORE AND' STEAM GENERATOR. THE.'STEADY STATE NATURAL CIRCULATION FLOW WAS CALCULATED AND COMPARED TO TRANSIENT NATURAL PREDICTIONS. THIS'COMIPARISON IS SHOWN IN FIGURE 2.* BOTH OF THE ABOVE COMPARISONS SHOW EXCELLENT AGREEMENT BETWEEN .THE ANALYTICAL VALUES AND NATURAL SUBROUTINE PREDICTIONS, THUS VERIFYING T.E PROGRAMM1ING. AND THE NUERICA SCHME OF THE.NATURAL SUBROUTINE..Y 3.3.1.2.. COMPARISON OF CADDS WITH. NATURAL TO DAVIS-BESSF. DATA* ~ ~ ~ ~ .* , I 11'4 TABLE 1. COM{PARISON OF NATURAL~ PREDICTIONS TOMANALYTICAL VALUES 'FOR A FLOW COASTDOWN Tmrl E(sec) Walycal o-NATUALK 1.0 .0.6452. .D.6454 2.0 ..0.4667 0 4669.3.0 0.3610' .0.3,611.. 400. 2917 0.2919 5.0..2432 0.2433 20. 0 K .25 0* 30.0 40.0*. ..50.0* .60.0: 70.0O* 90.~0 100.0*0.1274.0.08336.0.06077.0.04723 0. 0380-* 002*736*0.'02099 b.01688 0, 01401* 0.01192.0. 01033-0.009084 0.08341,* *. .06081.* .0.04726* 0.03832* ~.o0.2738.

0.-02101* 0.01689 0.01402 0.01193* 0.01034 0.009090~..r a .I t ..a I E

-... FIGURE l. PRIARY. SYSTEM SCHEMATIC........HOT LEG..e 1-.I* t I ..V -...I STEAM GENERATOR 4, L L'-I N, CORE/7-.-... .V 1'... ..t'I ..'. ':'FLOW 45 Lad... .................. .... ......... .....e...-fl.M. ...w ....I t ... .;. 1, *\,. -, d 1. ..:;W/w0$FIGURE 2. COMPARISONPOF NATURAL PREDICTIONS TO ANALYTICAL STEADY STATE NATURAL CIRCULATION FLOW. .: ,0.9 0.8 0.7 -4 0.6 0.5..I I-1 O&l ... I?XAIltia Natra *. .: ..-: Ciclto Value --. ,.I .:: I-0 -I.I.': 0.'4.I 0.3.I I I ..0.2'.8 I I ...Z I 0.11-_ .,..k I.: .I...i .I.....I--a C 0 10 20 30 .40 50 60 70. 80 90 100- 110 120......-% ---/m ..I -3.2.2. MODELLING ASSPTIONS...'. THE NATURAL CIRCULATION RESULTS. WHICH ARE SUBSEQUENTLY REPORTED WERE OBTAINED USING A ONE (1) LOOP CADDS/NATURAL MODEL. THE ASSUMPTION OF ONE LOOP MEANS THAT ALL DECAY HEAT REMOVAL OCCURS THROUGH ONE ACTIVE STEAM GENERATOR: AND'THAT' THE STEADY STATE NATURAL CIRCULATION! IN THE IDLE LOOP IS ZERO. THE CADDS MODEL ALSO ASSUM4ES'THAT ALL FLOW FROM -THE OPERATING PUMP GOES THROUGH THE-CORE PRIOR TO PUMP TRIP. WHILE-THIS ASSUMPTION IS NOT ENTIRELY ACCURATE UNDER-FORCED-FLOW CONDITIONS (BECAUSE'OF REVERSE FLOW -IN THE IDLE LOOP), IT DOES' NOT PREVENT THE OBTAINING OF A VALID STEADY STATE-NATURAL CIRCULATION SOLUTION UNDER CONDITIONS CONSIDERED HEREIIJ. -' ANOTHER ANALYSIS ASSUMPTION IS THAT THE PUTMP ROTOR LOCKS AT TIME OF TRIP AND .A LOCKED ROTORPUMP RESISTANCE IS USED FROM THEN ON. THIS RESULTS IN THE MAXIMUM FLOW COASTDOWN AND MAXUflUi' TRANSIENT CORE TEMPERATURE WHILE NATURAL-CIRCULATION IS BEING-ESTABLISHED. .THE PRINCIPAL VALUE USED FOR CORE FORM LOSS HAS BEEN CALCULATED FROM CORE.FLOW.AND .PRESSURE DROP INFERRED FROM PMI-2 HOT LEG FLOW: MEASUREMENTS. GIVEN THAT THE CORE 'PRESSURE DROP IS APPROXIMATELY.18 PSI AT A CORE FLOW OF ABOUT* 4500 lbm/sec AND THAT THE PRESSURE DROP IS ENTIRELY FORM LOSS, A FORM LOSS.FACTOR (K) OF 1100 CAN BE CALCU-LTED, WHICH IS ABOUT 200 TIMES THE NORMAL CORE FORE LOSS. OTHER CORE. FORM LOSS FACTORS-USED IN THE NATURAL CIRCULATION '. 1 ANALYSIS ARE 5 5 (NORMAL),'330 (60:TIES NORMAL), AND 5500 (1000 TIMES NORMAL).ADDITIONAL MODELLING ASSUMPTIONS CONCERN THE CORE BYPASS FLOW AND VENT VALVE FLOW. MOST OF THE ANALYSIS ASSUMES:A CORE BYPASS FLOW OF 4.6%, BUT ONE CASE CONSIDERS.A 30%'BYPASS FLOW. IN ALL CASES VENT'VALVE FLOWS ARE.NEGLECTED. I .--..: .1 '- .: -.... .... .. .- .53.2.. .ESULTS' IN ORDER TO INVESTIGATE THE-FEASIBILITY OF COOLING THE CORE WITH NATURAL CIRCULATION -USING WATER TO WATER IIEA~f TRANSFER IN THE STEAM GENERATOR, A STUDY WAS PERFORMED USING THE. FOLLOWING. MATRIX: CORE POWER -5 M4W FEEDWATER

TEMPERATURE

-1006F ONLY. A STEAM GENERATOR OPERATING CORE -PRESSURE DROP 60 TIMES NOMINAL~- FEEDWATER FLOW 1100, 3000,.5000 GPM~CORE BYPASS 5 THE STEADY STATE'.RESULTS OF THIS'STUDY.ARE SHOWN IN.TABLE 2.TABLE.2.--FEEDWATER FLOW CORE'INLET COEOUTLET PRIMARY. SYSTEIM GPM TEMPERATURE

F TEMPERATURE

`F 'FLOW LBM/SEC..1100 123

151 .210, 3000- 103 ... 18325~5000 101 116 338* FROM T-HE ABOVE TABLE, WE.SEE THAT IF TiE -FEEI)WATER FLOW EXCEEDS THE PRIMARY SYSTEM FLOW, THE COOLING WILL7OCCUR-HIGH ENOUGH IN THE OTSG TO DEVELOP REASONABLE
NATURAL..CIRCULATION FLOW RATES. BASED'ON ,THESE.RESULTS, A .FEEDWATER FLOW RATE OF 3000 Gpm AT ioboF'WAS CHOSEN AS THE MODE "FOR REMOVING THE DECAY HEATAN* 'FURTHER PARAMETER STUDIES -WERE PERFORMED..

TABLE 3 SHOWS THE STEADY STATE RESULTS FOR. THE CASES RU N. TABLE 3 CORE PRESSURE DROP %CORE CORE SYSTEM CASE CORE POWER LOSS CHARACTERISTRICS BYPASS INLET OUTLET FLOW NO. LEVEL MW X NOMINAL FLW.TMP F TEMP. *F LBHI/SEC.1 3 1 5 101.9 109.7. 380 2 3 60 5 101.3 111.1 ~ 305-33 .200 .109.:8 113.7 231.4 3 1005 100.4. 121.3 4 53200 30' 100.9 11. 235 6 ..2 .60 .5. 100. 9. 107.-7 .:292 CASE NUMBER FIVE IS THE BEST ESTI1HATE OF.THE EXPECTED CORE BEHAVIOR IF NATURAL CIRCULATiON IS INITIATED WHEN THE DECAY HEAT LEVEL IS 3: MW. A CORE PRESSURE. DROP.LOSS CHARACTERISTIC OF 200 TIMES14NOMINAL. IS SLIGHTLY.LAR'GER THAN THAT REQUIRED*TO-PRODUCE AGREEMENT-WITH CURRENT MEASUREMENTS OF' REACTOR; COOLANT -FLOW SPLITS.'(EE P. S. BARTELLS, "CR PRESSURE. DROP FOR NATURALJ CIRCULATION CALCULATION",*-APRIL 9, 1979).. THE BEST ESTIMATE FOR CORE BYPASS FLOW IS BETWEEN 22 AND) 27 PERCENT. (R. M.. GRIBBLE, "CORE BYPASS FLOW FOR CORE 'BLOCKED, AT TOP ONLY"', APRIL8, 1979.)-FIGURE 3 SHOWS A-PLOT OF CORE FLOW'VERSUS. TIME AND FIGURE 4 SHOWS.A PLOT. OF TEMPERATURE AT THE HOT LEG RESISTANCE THERMOMETER VERSUS-TIME FOR CASE FIVE.NOTE THE':TIME DELAY BETWEEN THE -START"OF'NATURAL -CIRCULATION ANqD THE.INCREASED HOT LEG TEMPERATURE. THIS IS DUE -TO THE LOW VELOCITIES .DURING -NATURAL CIRCU-LATION.THE STEADY.STATE .NA~TURAL 'CIRCULATION FLOW OF 235 LBM/SEC FOR CASE FIVE PROVIDES AMPLE COOLING FOR THE -CORE.: IN FACT, IF THE HOT:BUNDLE ONLY RECEIVED 10% OF--THE NOM1iINAL FLOW D'UE.TO BLOCKAGES, A CORE FLOW of 25LBM/SEC WOULD .BE ENOUGH TPRVENT BOILING-IN THE BLOCKED BUNDLE,(DA.FNSOTREUEDLW VERSUS COOLING EFFICIENCY", APRIL -10, 1979).' COMPARISON OF CASES THREE AND FIVE.SHOWS'THE.EFFECT OF CORE BYPASS FLOW ONi.SYSTEM FLOW AND CORE COOLANT TEMPERATURE INCREASE.' AS EXPECTED, THERE IS VERY LITTLE INFLUENCE ON SYSTEM FLOW.. AS THE CORE BYPASS FLOW INCREASES THE FLUID GOING THROUGH THE CORE IS HEATED MORE AND WHEN THE TWO STREAMS MEET -AND ARE MIXED THE TEMPERATURE IN THE UPPER PLENUM IS NEARLY THE SAME; THUS THE DENSITY DISTRIBUTIONS ARE NEARLY THE SAME FOR BOTH RUNS AND THUS-THE FLOWS ARE NEARLY EQUAL. SINCE THE'CORE.COOLANT TEMPERATURE.RISE.IS QUITE SMALL, SUBSTANTIAL BYPASS CAN OCCUR AND STILL HAVE ACCEPTABLE CORE COOLING.COMPARISON OF CASES TWO, THREE AND FOUR- SHOWS THE'EFFECT OF INCREASING THE EFFECTIVE CORE RESISTANCE TO FLOW.. AS.EXPECTEDj THE FLOW DECREASES AND THE TEMPERATURE 'RISE ACROSS THE CORE INCREASES; HOWEVER, EVEN FOR PRESSURE DROPS 1000 TIMES NO'SaNAL (ABOUT FIVE TIMES WHAT WE ESTIMATE) THE CORE FLOW IS ADE-QUATE TO COOL THE CORE. :'Z/..L.m ..w.4.. ())4 .1T=... L7 A.1 4 I ;Z.... .........LL______________________________3__A__ ! .......) .O., ....... .., .; --... v .. .., .................... .....!;' ' .;, 4 , :. _ ...... _ .:...... 4 too I x.0..I a,-000 -1 " 0 I I! ...-...T j A t R ( ef C ...*-.s *@9 z ..f .: .: .*....:.+@ A::* -.@ .@- ;g- g: ): z In &:;:.Ad) .......* ........* .: N ..Us_ , !A>D J .= :-,.,._--.....jU' C.".. .VT"-T Ourtl i.ff-AC.6IT.Oi.j; -_....~,_______ §.z::k ----------7Z -- --- --- '. ~ .~-9 I 7r_77-I 1 W 2c00 6f4m)J D).* .-0..~ COD i._- .+ TIP E .( re. C.)J s l O OI O a *.4 .m I.I. .a.I-.-1. Y. H. Hsii et.al., CADDS -Computer Application to Direct Iigital Simulation of Transients in PWRs with or without SCRAM, Revision ,-BAW-10098P, Rev. 1, becember 1977. -'.a.....: \., 1.:......................... ....... ...( %,': * .! .,...-............,... ....., ............: : .*-. J......... .....-I .. ' ' -;........., , * ..:., ..' ..... ....,.... ..... .. ...' .,., E..... .... ........i , ...,, s,; _:.....* ...: -.....w I ...I4 -....0 t. .. 14'~.3. 5 COMPARISON WITH ALTERNATE CALCULATIONS

ACOPUTER PROGRAM HAS BEEN WRITTEN TO PERFORM THE SAME' TYE OF STEADY STATE ANALYSIS AS HAS BEEN DONE: BY HAKD. THIS PROGRAM INCLUDES A 'FORCE BALANCE-ON.THE.VENT.VALVES.AND WILL TREAT THE RECIRCULATION FLOW IF.THE VENT 'VALVE.OPENS. THE PROGRAM ALSO CONSIDERS-THE IDLE LOOP AS.A FLOW PATH. FOR A RUN, WHERE.THE CONDITIONS.WERE.SIMILAR TO CASE 1 OF 'TABLE 3 (POWER- 3 MW, CORE PRESSURE'DROP AT ITS NOMINAL CON~DITION,`FEEDWATER FILOW OF 3000 GPM), THIS PROGRAM .COM1PUTED.

A CORE FLOW.OF 313 LBM/SEC WHERE CADDS/NATURAL PREDICTED.- 380.; THIS IS GOOD-AGREEMENT CONSIDERING THE 'SMALL DIFFERENCIES IN DENSITY.* 'AND PRESSURE 'DROP .WHICH OCCUR, AND THERE -WAS A;DIFFERENCE IN PRIMARY COOLANT TEMPERATURE WHICH-WOULD ACCOUNT FOR-PART OF.'THE DIFFERENCE. THE'STEADY 'STATE CODE PREDICTED CORE AT'OF 9 DEGREES~ VRUS CADDSNARLSPEDCINO 7.8. THE STEAD STATE CODE COMPUTED THAT THE PRESSURE DIFFERENCE.WAS NOT SUFFICIENT TO OPEN THE-VENT VALVES AND THE IDLE LOOP 'FLOW 'WAS ONLY 32 LBS;THUS, THE SIIIPLIFICATIONS USED IN-~THE CADDS'-NATURAL-PROGRAM ARE REASONABLiE.- 3...1.2. COMPARISON OF CADDS WITH NATURAL CIRCULATION CALCULATION TO DAVIS-BESSE EXPERIMENT IN ORDER TOVERIFY THE VALIDITY OF THE CADDS/NATURAL CALCULATIONAL RESULTS, A SPECIAL. CASE -SIMULA.TING THE DAVIS-BESSE. PLANT WAS RUN -AND THE RESULTS COMPARED-WITH THE REPORTED EXPERIMENTAL MEASUREMENTS. THE FOLLOWING'TEST-CONDITIONS WERE USED IN THE.CADDS/NATURAL INPUT.1) 2-LOOP.CONFIGURATION

2) FEEDWATER FRO, ~THE AUXILIARY FEEDWATER SYSTEM* 3) REACTOR POW4ER WAS CONSTANT-AT 3.85% OF 2772 M4Wth 4) TWO DIFFERENT FEEDWATER RATES. (46 AND 460 LBM/SEC) WERE USED.THE RESULTS AESHOWN IN FIG. ____ SSONB H IUE THE SYSTEM FLOW RATES DO NOT'REALLY STABILIZE IN 1800 SEC AND:SMALL OSCILLA-TOSSIL CAN BE SEEN; HOWEVER, IT APPEARSTHAT.

THE PRIlAYSSE FLOW.) RATES ARE APPROACHING A STEADY STATE NATURAL CIRCULATION, FLOW:RATE OF 6.0% FOR TH O EDAE ATE AND 4.4% FOR.THE HIGHER FEEDWATER RATE.THE REASON FOR THIS IS THAT.THE .HIGHER FEEDWATER.RESULTS. IN LOW4ER PRIARY SYSEM EMPRATREWHCH IMPLIES HIGHER DENSITY'.AND VISCOSITAN CONSEQUENTLY, HIGHER FRICTION.THE DAVIS-BESSE EXPEkIHENT REPORTED THAT THE NATURAL CIRCULATION

FLOW RATE RAkNGES FROM 4.6 TO'5.1%, DEPENDING ON THE WATER LEVEL OF THE STEAM GENERATOR.:

IT APPEARS THAT THE CADDS/NATURAL PROGRAM W4ITH THE TWO EXTREMES IN 'FEEDWATER FLOW PREDICTS'NATURAL CIRCULATION FROM'RATES THAT BRACKET THE'EXPERIMENTAL RESULTS. THE--DIFFICULTY IN COMIPARING CADDS/NATURAL DIRECTLY TO THE DAVIS-BESSE DATA IS DUE'TO THE FOLLOWIN1G:

1)' THE OPERATOR VARIED AUXILIARY FEED PUMP SPEEDS SO MUCH THAT THE 'FLOW RATES RANGED FROM 200 TO 1400 GPM4 AND'THAT.

NO AvERA(E STADY STATEFLWRT COULD BE DISCERNED FROM THE'DATA. --2 -- --.i... .m 2) DURING THE EXPERIMENT, THE MAIN FEEDWATER WAS TURNED OFF AND ONLY THE AUXILIARY FEEDWATER WAS USED. AT PRESENT, CADDS DOES NOT HAVE THE CAPABILITY OF MODELLING THE AUXILIARY FEEDWATER PROPERLY.'0,:-.I ..i 11 ...I m .4.W. . 11~~yv~,.vq't.G. A. MEYER fl_ ,Jt Vt _ , APRIL 10, 1979 TIME: 1630 3.7 CORE THERMAL BEHAVIOR A. ANALYSIS OF CURRENTOPERATING CONDITION CORE BLOCKAGE STUDY TYO METHODS HAVE BEEN'USED TO EST I TE'THE EXTENT OF THE CORE BLOCKAGE. THE FIRST METHOD INVOLVES THE USE' OF THE INCORE THERMOCOUPLES'TO DETERMINE THE CORE'OUTLET 'TEMPERATURE. THIS METHOD PREDICTS A CORE FLOW OF LESS THAN 1 X i0 6 LB/HR.1 THE SECOND METHOD INVOLVES MATCHING THE PUMP CODE TO THE DATA DU I THE PLANT. THE PUMP CODE CALCULATES FLOW SPLITS FOR DIFFERENT'PUMP CONFIGURATIONS. THE CODE WAS MODELLID TO MATCH THE PRE-INCIDENT CONDITIONS. THE CORE RESISTANCE WAS INCREASED UNTIL THE CODE PREDICTED THE SAME HOT LEG FLOWS THAT ARE'BEING MEASURED BY THE TWO GENTILLE DELTA P'S.6 THIS CALCULATION PREDICTS A CORE FLOW OF 13 x 10 LBM/HR AND A CORE PRESSURE DROP OF 18 PSI. THESE CONDITIONS. ARE CALCULATED WITH A CORE RESISTANCE APPROXIMATELY 200 TIMES THE NORMAL RESISTANCE 2'3'8 OR A FORMLOSS RESISTANCE OF 1650.TO DETERMINE IF A CORE RESISTANCE O0 THAT MAGNITUDE IS FEASIBLE AN ESTIMATE OF THE CORE DAMAGE WAS DEVELOPED. A CURRENT ESTIMATE OF THE MATERIAL AVAILABLE FOR CORE B3LOCKAGE IS :

.G. A. MEYER* APRIL 10, 1979 TIME: 1630_ '

63 FT 3 U2' 94 FT 3 U2 OTHER -10 FT 3..167 FT 3 'THIS DEBRIS IF SPREAD EVENLY ACROSS THE CORE. COULD FORM A PACKED BED) OF DEBRIS THREE FEET THICK.4 THE RESISTANCE (FORMLOSS COEFFICIENT)

FOR THIS BED HAS BEEN CALCULATED AT APPROXIMATELY 17005 WHHICH IS IN GOOD AGREEMENT WITH THE TOTAL CORE RESISTANCE CALCULATED USING THE MEASURED FLOW SPLITS. THE METHOD TWO CALCULATION OF A ONE PUMP CORE FLOW OF 13 x -6 LB/HR AND THE ASSOCIATED BLOCKAGE IS THE BEST ESTIMATE OF THE CURRENT CORE CONDITIONS FOR THE FOLLOWING REASONS.1) CALCULATIONS HAVE SHOWN THAT SUFFICIENT UO 2 COULD.BE "PACKED" AROUND THE THERMOCOUPLE TO GET A TEMPERATURE READING 350 HIGHER THAN THE FUEL ,6 ASSEMBLY BULK EXIT TEMIPERATURE. 6 2) CALCULATION INDICATE THAT ON THE AVERAGE THE THERMO- -COUPLE'WERE READING 5 0 F TOO HIGH BEFORE THE I X T. 9 3) IT SEEMS IMPROBABLE THAT SUFFICIENT BLOCKAGE COULD EXIST TO REDUCE THE CORE FLOW TO THE 1 10 LB/HR PREDICTED IN METHOD ONE.4) THE SHIFT IN THE THERMOCOUPLE READING BEFORE, DURING, AND AFTER THE SWITCH FROM PUM\IP Al TO A2 INDICATE THAT THE CHANGE IN THE FLOW DISTRIBUTION IS CAUSING 4)G. A. MEYER APRIL 10, 1979.TIME: 1630 THE DEBRIS TO SHIFT WHICH IS-AFFECTING THE THERMOCOUPLE READINGS. THE SHIFTING DEBRIS MAKE THE METHOD ONE ANALYSIS MORE SUSPECT AND LENDS CREDENCE TO THE HYPOTHESIZED LARGE BED OF DEBRIS CALCULATED IN METHOD TWO... 0 ....-- : -:- -L 2215 --. .REFERENCES X.1. JH Jones to GA Meyer "Estimate 6f Core Flow".'2. PS Bartells to Distribution, "Core Pressure Drop for Natural Circulation Calculation", 4/9/79 3. .x. Hiatt to. G. A. Meyer, "LYNXL Model for TMI-2 Blockage Study", 4/10/79 4. Core Condition Task Force, "Estimate of Loose Core Debris Volume (4/9/79 -2000),-4/9/79. .5. P.' J. Henningson to G. A. Meyer, ."I)amage Model -Fluidized Bed", 4/10/79 6. P. J. Henningson to C. A. Meyer, "Increased T.C. Readings Due to Proximitty of Fuel Particulates" 7) T. L. Wilson to J. T. Willse, "Discripancy between Thermocouples and Outlet RTD Temperature Measurements", 4/9/79..*8) R. M. Gribble to J. D. Carlton, "Core Bypass Flow For Core Blocked at Top Only",: 4/8/79* 9) J. A. Weimer to G. A. Meyer, "Incore Thermocouple Error Evaluation", 4/10/79.. ...-., ..-......-.-OTHER INFORMATION .....1) R. tI. Hiatt to G. A. Meyer, "Core.> ..Operation". 4/10/79.. ..:.: ..: ............ ..........................-.', -........................:.: ' ' ' :...Flow Distribution for One Pump and two Pump C E -c)-- IrORE THERMOCOUPLE --._.DESCRIPTION The incore thermocouple design for TMI-2 is a grounded junction chromel-Alumel detector The-vt-afion of the detector in relation to the fuel assembly upper end fitting is shown on Figure 1. The thermocouple wire diameters are approximately -10 mils, the sheath is 62 mils OD Inconel tubing, and the insulation material is A1 2 0 3.The location of the thermocouple within the instrument tube is shotn in Figure 2. There are 50 thermocouples at locations in the core shown on Figure 6.Ji-THERMOCOUPLE READINGS: The highest temperature readings of the thermocouples are plotted from 0400, 3/28 to 1300, 4/6 on Figure 3.pt about 1300 on 4/6 a RC pump switch from 1A to 2A octui~eTich caused a redistribution of thenrnoouple readings. The change in thermocouple readings as a result of the pump switch is shown on Figure 4. The highest thermocouple readings after the-pump switch are shown on Figure* 5. Complete sets of thermocouple readings at selected 'times of each day are given on-Figure 6.. * .The major point addressed in this section is to evaluate the thermocouple readings to determine the validity of the temperatures. This was done by examining detector accuracy and by evaluating fuel pellet debris accumulated around the thermocouple. The results of this evaluation are that the thermocouples are reasonably accurate and that the high temperatures are a result of fuel particles in the upper end fitting. .. .~ DETECTOR ACCURACY.*.PAST EXPERIENCE WITH THERMOCOUPLES INSTALLED IN B &W PLANTS SHOW TlWO1 BA~SIC TRAITS.'.THEY TRACK FAIRLY WITHOUT- ANY OBVIOUS CHANGE IN CORE.'CONDITIONS. THERMOCOUPLE READING CHANGES GREATER-THEN 2 F ARE REFLECTED 'IN OTHER MEASUREABLE .LOCAL.OR CORE* CHANGES'.*THE THERMOCOUPLES WERE EXPOSED TO TEMPERATURES.GREATER THAN (1800 F) DURING THE INCIDENT AT TMI- 2.TO DETERMINE IF THESE ELEVATED.TEMPERATUrRE$IWOULD PRODUCE INACCURATE THER-MOCOUPLE READINGS, -ESTS WERE CONDUCTED) AT THE BA'LFAD SUBSIDIARY OF BWq.ITHESE TESTS, .4. *THEI!OCOUPLES .ERE RAISED TO.' -2000- F FOR FOUR HOURS. THE THERMOCOUPLE READINGS WIERE THEN CHECKED AGAINNST' KNOWN READINGS OVER THE'R ANIG E OF 20 0 T7O 1000 TF. ALL TEST. THERMOCOUPLES READ WITHIN STEDYTATOF THE CALIBRATION VALUECAI WAS.IHI .E. IT WA .OCU THAT THE flIGH TEMPERATURE EXPOSURE Wl;OULD D NGORT SIGNIFCANTL ALTER THE THERlOCOUPL E READINGG.* , THERiOCOUPLE DATA FRIOM TI.-2 DURING NORMAL OPERAT° ON BEFORE THE LOFW TRANSIENT WAS EiXAMINED TO DETERMINRE-IF A SYSTEIATIC BIAS EXISTED. NO DATA WAS AVAILABLE AT ZERO POWER CONDITIONS, THEREFOREDATA FROM SEVERAL-POWER LEVELS WAS EXTRAPOLATED DOIW JO ZERO POWER. THE-RESULTS OF THIS ANALYSIS SHOW THAT A + 5°F BIAS IS POSSIBLE.:.. ON.THE THERmOCOUPLE READINGS.- ....THE POSSIILITY O)F DE-CALIBRATION SIN'CE TIE INCIDENT I .AS.ALSO BEEN .EVALUATED. SIX THERMOCOUPLES WHICH HAD A* .TEMPERATURE RISE OF 7 to 33 0 F OVER A 5 DAY PERIOD--W -ERE EXAMINED. NO MECHANISM COULD BE POSTULATED WHICH WOUJLD CAUSE DECALIE fTION TO THE EXTENT SHOWN BY. THE.DATA. -IT W1ifAS CONCLUDED THAT THE THERIMOCOUPLE READINGS WERE ACCURATE AND THAT TRUE TEMPERATURES IWERE BEING-.MONITORED TO

5°F. --:* THE RESULTS OF THE THERMOCOUPLE TESTS AIN-D EVALUATION'S INDICATE THAT THE TEMPERATURE READINGS ARE GENERALLY ACCURATE WITH A POSSIBLE + 5°F BIAS.

3H9P(3)-HEAT CONDUCTION FROM FUEL PELLET DEBRIS An evaluation was performed to determine if the quasi steady'thermocouple readings of 100 to 240 F above coolant temperature could be caused by fuel'pellet agglomerates located on spacer grids or in' end fittings. -One evaluation was performed assumi ng al l the pellets from one Trid span were:. ce0ss teu olt caqt .hcQt is qelnnraif evenly distributed, in this mass to produce boiling in the annulus between the instrument string.and the' guide tube. The steam would be vented out of the guide tube below the upper end fitting. Heat transfer to the thermocouple would.be from axial conduction along the instrument string. ..Due to the large amounts of fuel debris required and the fact'that the instrument guide tube would be cooled above the fuel mass, it is concluded that this mechanism is not the most likely reason for the high thermocouple readings.It is more likely that fuel pellets and pellet fragments have collected .in the upper end fitting and/or the mixing cup. Reactor coolant flow is sufficient to carry fuel pellet fragments above the upper end fitting where .they would settle out in this stagnent region. An analysis was performed which assumed various* ..' ~cc .' .'.widths of fuel assumulation in the upper end fitting around and in the' mixing cup. The results of this analysis is shown on Figures .7 and 8. The case.with fuel.particles inside the mixing cup (T.C. Well) shows a small ( 10 0 F) increase in temperature which is insignificant. Fuel debris to'a width of 3 to 4 inches in the upper end fitting itself. does produce the magnitude atAT's which are seen on.'the readings. This amount of fuel could easily fit within the upper end fitting whi ch has an interiour width of 7 inches. -Bibliograph for 3.7A(2) and (3) For'A.B. Jackson-3217 1. Memo, T.L. Wilson to E. Oelkers, "Reliability and Uncertainty of Thermocouples Following Loss of Feedviater Transient", April 5, 1979.2, Memo, J.A. Weimer to G.A. Meyer, "Response to Thermocouple Request", April 5, 1979.3. Memo, J.A. Weimer to G.A. Meyer, "Incore Thermocouple Error Evaluationi',** April 10, 1979.4. Memo, T.L. WVilson to J.T. Willse, "Discrepancy Between Thermocouples and Outlet RTD Temperature Measurements", April 9, 1979.5. Memo, J.T. Willse to Engineering Operations Manager, "Action Item 143", April 6, 1979.6. Memo, P.J. Henningson to G.A. Meyer "Increased T.C. Readings Due to Proximity of Fuel Particulates", April 10, 1979.., . 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THEVrADcocKs Et WILCOX CO>...~~~~~~- --I- -.I ........ ....................... .. ...*. TtZ---- a.r~su.IKQ -wRE- 1 S P-ND--sTTNG NUMBE RS AND LOCATIONS 177 FA CORE: F-F7: _L 1 LII STRIGt NO...r.-, _9"'tI 30 A013-f I ....3 6 3 (J._3 I f I --V 1G t..x ..6, , ..................... .. ,........ -,._.'; q Agq 11*5.r P6 1~3 Ii 3 2i* I -/ .1 ? __ --Z Qol k3l -.1__ __r _ _ f t _.I_-rtL .f.~ I t _.1. -I2_ -21.rglq6 A0o, 3793J'1 42______t l 30 l I I _____1 ___ _I ,fna. --__ _ o/ H4 j jj f____i .S cir(0 46" 9Lq* It)13 Ito, 2A /0 I 7-9/a c10 ji __j f;'i..f __ __1I _ _ _ _ _ _ _ _ _ -._ _ I I Pq .I1 -A-- .-. 4, ... I W 4 8I di 9Z P40o()'927 Ca ..,n.a0 N .4 -1.. --/,C1061-~ ...... f ,--- -93s 40f'l0_ Wi I5 bs_ .1 .__'^_^D., _l fi 2 -3 q S-6 7 8 7 *0 11 -. I /3 I'f 1_P.4 r. .. --.---T---At. ..L )I-fia- -TH 1E tJABCOCI' & WILCOX CO.: -*:~~wc ..- B IC'GS_, ...1 S~LD--SitG -NUMBE RS AND [1T 1 ISTi~i6 "O.* ~ LOCAT IONS ~- 177 FA CORE -* [L-'~ ~ .,.,, .31 AW/30.. 1/3?.--- ----if -'T'I. I __. .iLZj~.___ I___ ___I36 J6965 lZro 2G1*y n i 35c ' u 6 i- 4'4 ---- --- .n 1 c2lo : j.g3s, ___ 38* t .a Wl.t, ,. t~l.r. _ o ,.3f 39gfi I~ { So 8.' 01{ -___ I ^^___ -,-8 ----. !- --~ -. ~ I J3 3 I I -I'S-. * -. ~ r '.-l -* I -i 4 ...fl I aa ..... ¶ 1 tI?. , 44A -.q?- I i3 ,. 4, Fl.; S.?S U 6 -A j -/44"9 6&__------ Wf, *--_-,_,_-,_,_ q5 I V..3I.6 -7 s a-./2 / /RaL e .i12 /13 : /, I i ?.7 t O /') (If? nIM 1 4 0 ..----. --.-TrHE rJAECOCK Wit-ILCOX CO. -t5).4 .. ..*.\5C ~ ' .-.SPN ..R~N NUBR AND'- ...., ,'- .....' '-, SPNQ D ST RI1, NXG NUM vlBE RS AN D...LNS .177 FA CORE*F .YI 7!..K. STRING NO-.--4 -.---- --I .....31 30 wI I _ _ T -- ------- ---3?..Z"i7.bit),g 5,1-t -.Th.n.. I LI6_____ I Bq3/49_ __ IrI .* 3 t *.l&w __ __ l3i I2) lI/ l G m _ .-.. .---i-- --i~~~*~- -- i I 201'1l- d0 .} ; f'. 31,~t~ .D }. -f l -AS 4 i~ ..-I' I K ' ....!.. 20J __._i-:- : -.-- -. --B- --9 YI __ '__ _ a _ _ *1,,-.*.V ----'.I'3.16.2,.. ?L: __ -o___ , _2 ___'..47.a..* C 4 .-.. * :- 1 > / ..i 2 3 9-I I PS,8 ,_. -- 5I-i--I._ _..3 j i .7 I*O -- 1/ /2 /3 1 -i..5 '.- ..~ '7 ..: _ -- --- ------- --- --Ttv ____..THE JjADCOCK & WILCOX CO. A a '4A*2: -. : --_ ..SPND 1 IRbNU NUlMBE R5 AND l LIATi ONS l-.I77 FA CORE -1 !I I-; --...31 AqA.3q sA5 I-___ ,, .., Si_ r .... -, , ...,,. ., -. ,.__U0 AW-C F F ., 4 .1 F I.

I I___ I *1___ ___ .I___ ___LJUL --.___ I I L.L.J [315 .L....I 951 50]F. I is 1(3 I'S%C, 5.16 ,oil ZA0 VIrt Z?/91.Al;r,Yl 4 2/31 I I 1* £1?.. 4 a S.-' --C-----. .1-.-- V *?-$ ?c;20,7.. 117 J/a A 0nL J. .4. -; -'Ij I...9fuf I 6 A 6f .4oe ,8.j 9rk/^/ C- ..I I.I---.....I , ' .2 -3 q .5-b6' 71I 7 8 .79 , I0 il I,?/3 1', ljlaI5
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I ________ _____________ I ..* I I... .............-..-. ........ .... ..._ _ _ _ 77:1O '.7 -_:_ _7_ _7__ _ _ _ _ _ _..IIC/I I ..-. .. ....I ............. ........ .. .. ..... ..L: hU47 S/izt4D1.$ .~::_7' -------. .I I. 7. ....I ....I ....! ....-I- -I -.,-.--:-I: ..'- I ;. ..-.-..... .-.. .I I I 1:::: 1: .,t ? ..- ...:.: --.- .....-. : I -, , i .. -.-.. -, , , '. f % -, , ,' , I .. -: .-- ., -, --I -: .--, '...- I.. -.!, -. ..-I ..- ....(...0 N.'-I 1.0 0 0 U-, 22-I., I--J Cu-I..Cu IA!0--I-- --------= :17: ..... ... ...... ............ .. .... ......7- =. ......f- .* -.......U z ~ zI ......j. .....____ __ i--~---T .---------.- 7.---: ____ ___ U._ __ ___ __e____ _______ ______~0~-(>:Z* .7.. .I.. ........-.I.4 ........7 7 ---. --.-4-~i ~ r -I.-.__._._._.__._._._._ .I -7.... .. ... .........z: 1 ~ ~ .4 ..L .~ ... ..-...........~5...___ _......__ __. .... .-..r.. *..: _ _ __ _ _ _ __ _ _JK.. ~ j.. ...4-. -... .... . --- /1O0/797-7;50 PM 3.7 B TH-ERMAL-HYDRAULIC EVALUATION OF NATURAL CIRCULATION' OPERATION OF THE TMI-2 CORE WITH NATURAL CIRCULATION HAS BEEN EVALUATED AND JUDGED ACCEPTABLE ON A LONG TERM COOLING BASIS.CORE MINIMUM FLOW REQUIREMENTS AS A FUNCTION OF TIME AND OPERATING CONDITIONS HAVE BEEN. DETERMINED. CRITERIA HAVE BEEN-ESTABLISHED FOR THEIACCEPTABILITY AND ESTABLISHMENT OF NATURAL CIRCULATION. THE EFFECTS-OF LOCALIZED HEATING HAVE BEEN EVALUATED AND LIMITS HAVE BEEN ESTABLISHED FOR THE INCORE THERMOCOUPLES AND-HOT LEG RTD'S.THE MINIMUM CORE FLOW-REQUIREMENTS FOR LONG TERUM COOLING ARE GIVEN IN REFERENCE

1. THE BASIS FOR THIS MINIMUM REQUIRED FLOW-IS NO BULK COOLANT TEIMPERATURE ABOVE SATURATION TEMPERATURE.

THE CURVES PRESENTED IN REFERENCE 1 INCLUDE A SAFETY MARGIN OF 300%- AND ARE COMPLETELY BOUNDED BY PREDICTIONS OF ACTUAL NATURAL'CIRCULATION CORE FLOW ASSUM4ING CORE RESISTANCES 200 TO 1000 TIMES THE RESISTANCE OF A NORMAL 1 77 CORE.-CRITERIA FOR THE-INCORE.THERMOCOUPLES AND HOT LEG RTD'S ARE DESCRIBED IN REFERENCE

3. THESE CRITERIA ARE DESIGNED TO ALLOW:ADEQUATE TIME FOR THE ESTABLISHMENT OF NATURAL CIRCULATION FLOW WHILE STILL 'ALLOW\ING FOR ADEQUATE SAFETY MARGIN AND TIME FOR TRANSITION TO.ONE OF THE ALTERNATIVE COOLING SYSTEMS. THE DEGREE OF CONSERVATISM IN THE CRITERIA 'IS DEMONSTRATED BY'ANIALYSTS RESULTS PRESENTED IN REFERENCES 2, 3,' & 4. THE REFERENCE 2 RESULTS DEMOSRAETHTTME N CES OF ONE HlOUR ARE REQUIRED TO I-IEAT THlE W1ATER CONTAINED ONLY IN THE CORE REGION TO SATURATION UNDER

-:.--- -.-* -- ----- ---.4/10/ 79 Page. 2 NO-FLOW CONDI*TIONS AND POWER LEVELS BELOW 3 MWT. REFERENCE 3 DEMONSTRATES THAT AN ADDITIONAL HOUR WOULD BE REQUIRED (@3 MWVT)TO BOIL OUT COOLANT FROM THE OUTLET NOZZLE LEVEL TO THE TOP OF THE CORE. REFERENCE 4 DEMONSTRATES THAT MORE THAN ONE HOUR IS-REQUIRED.(@3 MWT) FOR-THE ADIABATIC HEAT-UP OF THE CORE FROM 200 0 F TO.2000° 0ASSUMING INSTANTANEOUS CORE UNCOVERAGE.. NATURAL CIRCULATION RESPONDS TO HIGHER FLOW RESISTANCE BY.-GOING FURTHER INTO TWO-PHASE FLOW UNTIL THE VAPORIZED DENSITY CHANGE IS SUFFICIENT TO OVERCOME SYSTEM RESISTANCE. WITH DEBRIS BLOCKING A GRID, FLOW WILL PASS.LATERALLY AROUND THE BLOCKAGE AND ADEQUATELY IMMERSE THE FUEL BEFORE AND AFTER THE BLOCKAGE. IF THE BLOCKAGE DEBRIS ALSO CONTAINS FUEL, THE" MAXIMUM. TEMIPERATURE AT THE BLOCKAGE CENTER WILL BE CONI)UCTIONLIMITED AND RELATIVELY INSENSITIVE TO THE MODE OF SURFACE HEAT TRANSFER. THE FLOW PATTERNS IN NATURAL CIRCULATION ARE DIRECTLY RESPONSIVE TO RESTRICTIONS, AND MAY. WELL SHOW LOCAL TWO-PHASE TRANSITIONS.AND RELATIVELY HIGH VELOCITIES AT PARTIALLY-BLOCKED HOTTER REGIONS.REDISTRIBUTION OF CORE DEBRIS IS ANTICIPATED WHEN THE PUMPED: FLOW IS TERMINATED. CHANGES IN THERMOCOUPLE INCORE TEMPERATURE DISTRIBUTIONS SHOULD BE EXPECTED. NOT ALL THERMOCOUPLE READINGS CAN BE EXPLAINED BY HYDRAULIC PHENOMENA. SONIE THERMOCOUPLES, PARTICULARLY NEAR THE CENTER, ARE CURRENTLY INDICATING LOCALIZED HEATING EFFECTS AND ARE NOT MEASURING BULK FLUID TEMPERATURES. HENCE, IT IS NOT REALISTIC TO REQUIRE.ALL INCORE THERIOCOUPLE MEASUREMENT BE BELOW SATURATION TEMPERATURE, NOR IS IT NECESSARY.

. ---.4/10/79 7:50 pm Page 3 REFERENCE 5 DISCUSSES THE ADVANTAGES AND DISADVANTAGES OF CORE COOLING AND NATURAL CIRCULATION AT VARIOUS PRESSURES.

IT IS DESIRABLE 04U PRUDENT TO At- k MP4SH4NIA1YDlOD1 lQMI-: MAINTAIN LONG TERM N COOLING AND NATURAL CIRCULATION AT-5g PRESSURESid THE *61- 6oo- 'Goc PSI As_ .

REFERENCES:

1) MEMO, GA. MEYER TO J.D. CARLTON, PLANT-DESIGN, "MINIMUM CORE FLOW-LONG TERM COOLING", APRIL 4, 1979.2) MEMO, L.L. LOSH/J.F.

BURROW TO G.A. MEYER THERMAL-'HYDRAULICS,"CRITERIA DURING ESTABLISHMENT OF NATURAL CIRCULATION", APRIL 10, 1979.3) MEMO, J.A. WEIMER, R.L. HARNE TO DISTRIBUTION, ;BOILING CONDITIONS IN CORE", APRIL 1, 1979.4) MEMO, J.H. JONES TO G.A.-MEYER, THERMNTAL-HYDRAULICS,"ADIABATIC CORE HEATUP RATES", APRIL 10, 1979.5) MIEMO, J.R. GLO0UDEI'IANS TO G.A. lEYER, THERMAL-HYDRAULICS,"COOLDOWN. PRESSURE", APRIL 10, 1979.I , , ..-, -G. A. MEYER-APRIL 10, 1979 TIME: 1520 3.7 CORE THERMAL BEHAVIOR C;. ANALYSIS OF ALTERNATIVES THE PLANNED MODE OF OPERATION, DISCUSSED IN SUB-SECTION B, IS TO ESTABLISH NATURAL CIRCULATION FOR THELONG-TERMN, COOLDOWN. IF THE SYSTEM RESPONSE AFTER THE REACTOR COOLANT PUMP IS TRIPPED INDICATES THAT'NATURAL CIRCULATION HAS NOT BEEN ESTABLISHED, THEN'THE FOLLOWING ALTERNATIVE'. MODES OF OPERATION WOULD BE CONSIDERED:

1) RESTART ONE REACTOR COOLANT PUMP'THIS WOULD RESULT IN CORE CONDITIONS ESSENTIALLY THE SANIE AS THOSE EXISTING NOW AND DISCUSSED IN SECTION A.IT IS ANTICIPATED THAI' SOME SHIFTING OF DEBRIS WITHIN THE CORE .1OULD'OCCUR, RESULTING IN A NUMBER OF FAIRLY RAPID CHANGES IN INCORE THERMOCOUPLE READINGS, SIMILAR TO THE CHANGES WHICH OCCURRED ON APRIL 6, i979 WIHEN THE Al PUMP TRIPPED AND THE A2 PUMP IWAS STARTED.' 2) DECAY HEAT SYSTEM (OR MODIFIED' DECAY HEAT SYSTEM)THE COOLING FLOW RATE AVAILABLE FROM THIS SYSTEM IS SIMILAR IN MAGNITUDE TO THAT AVAILABLE'WHEN NATURAL' CIRCULATION IS 'SUCCESSFULLY ESTABLISHED, THEREFORE THE ANALYSES OF SUBSECTION B APPLY. THE MINIMUM REQUIRED'LOWRATE, PROVIDED IN REFERENCE (C.1) INCLUDES A FACTOR OF 5.8 ON THE FLOWV CALCULATED AS THAT REQUIRED TO AVOID BULK-BOILING WITHIN THE CORE. LOCALIZED BOILING IS EXPECTED TO OCCUR, AND THIS COULD CAUSE SON1E OF THE INCORE THERMOCOUPLES, SUCH AS THOSE LOCATED IN

-- ---.G. A, MEYER... -APRIL 10, 1979.TIME: 1520 I CORE LOCATIONS H8 AND HS, TO INDICATE TEMPERATURES SOMEWHAT HIGHER THAN THE AVERAGE AND POTENTIALLY HIGHER THAN THE SATURATION TEMPERATURE (ASSUMING THAT THESE THERMOCOUPLES ARE IN CONTACT WITH AGGLOMERATIONS OF FUEL FRAGMENTS).. THE FACTOR OF 5.8 ACCOUNTS FOR A 27%. CORE BYPASS FLOW RATE WHICH, IN TURN, REFLECTS-AN EXTREMELY HIGH CORE FLOW RESISTANCE RESULTING FROM THE POSTUALTED CORE BLOCKAGE.3) HIGH PRESSURE INJECTION SYSTEM THE MINIMUM FLOW RATE SPECIFIED FOR THIS MODE OF OPERATION-. IS BASED UPON THE SAME EVALUATION AS THAT FOR THE DECAY HEAT SYSTEM, ANI) REPRESENTS THAT FLOW RATE NECESSARY TO AVOID BULK BOILING. THE AVOIDANCE OF BULK BOILING, ALTHOUGH NOT CONSIDERED TO -BE ABSOLUTELY NECESSARY, IS DESIRABLE TO MINIMIZE HOT SPOT TEMPERATURES IN CORE LOCATIONS SUBJECTED TO SEVERE BLOCKAGE. THE FLOW-REQUIREMENT SPECIFIED IS CONSERVATIVELY BASED UPON A DECAY HEAT'VALUE OF 4 MWT. FOR LOWER POWER LEVELS THE MINIMUM REQUIRED FLOW DECREASES IN DIRECT PROPORTION TO THE DECAY HEAT LEVEL. -REFERENCE C.1 A. B. JACKSON TO C. C. ENGLAND, "REQUIRED FLOW FOR CORE COOLING," APRIL 10, 1979. - THE BABCOCK & WILCOX COMPANY POWER GENERATION GROUP .EL I M / /C. A. MEYER o L L. L. LOSH/J. F. BURROW S' DS 663.5 Cust. File No.TMI-2 ..or Ref.Subj- CRITERIA DURING ESTABLISHMENT OF NATURAL CIRCULATION Date 5 40 PM APRIL 10, 1979 this l.tter to cow.r one customer end one svbjiecl eny.THE TRANSITION FROM FORCED CORE COOLING USING ONE REACTOR COOLANIT PUMP TO NATURAL CIRCULATION WILL BE ACCOMPLISHED OVER A PERIOD OF TIME. HEATUP OF THE CORE'COOLANT WILL GENERATE A DENSITY GRADIENT AXIALLY IN THE CORE WHICH WILL PRODUCE THE DRIVING FORCE FOR THE CIRCULATION FLOW. THE TIME REQUIRED TO ESTABLISH NATURAL CIRCULATION IS THEN DIRECTLY RELATED TO THE HEATUP RATE OF THE CORE WHICH GENERATES THE DRIVING FORCE. TO ESTIMATE THIS HEATUP TIME THE FOLLOWING ANALYSIS WAS PERFORMED: (1) IT WAS ASSUMED THAT THERE WAS NO CORE FLOW, (2) THE ENTIRE CORE COOLANT INVENTORY WAS RAISED FROM T TO TSA (3) NO NET STEAM QUALITY WAS-IN. SAT'GENERATED, (4) CALCULATIONS WERE BASED ON GROSS CORE AVERAGE CONDITIONS. FIGURE 1 DEPICTS THE CORE DECAY HEAT GENERATION AS A FUNCTION OF TIME. -THE RESULTS OF THIS ANALYSIS ARE SHOWN IN FIGURE 2-4 FOR VARIOUS ASSUMPTIONS ON -T SYSTEM PRESSURE AND CORE POWER (TIME AFTER SHUTDOWN). THESE CURVES WERE USED TO ESTABLISH A MAXIMUM TIME REQUIRED TO ESTABLISH NATURAL CIRCULATION. NATURAL,_CIRCULATION RESPONDS TO HIGHER FLOW RESISTANCE BY GOING FURTHER INTO TWO-PHASE FLOW UNTIL THE VAPORIZED DENSITY CHANGE IS SUFFICIENT TO OVERCOME* SYSTEM RESISTANCE. WITH DEBRIS BLOCKING A GRID, FLOW WILL PASS:LATERALLY AROUND THE BLOCKAGE AND ADEQUATELY IMMERSE THE FUEL BEFORE AND AFTER THE BLOCKAGE.IF THE BLOCKAGE DEBRIS ALSO CONTAINS FUEL, THE ITAXIMU1-1 TEMPERATURE AT THE BLOCKAGE CENTER WILL BE CONDUCTION LIMITED AND RELATIVELY INSENSITIVE TO THE 5:40 P.M.G. 'A.'MEYER APRIL 10, 1979* MODE OF SURFACE HEAT TRANSFER. THE FLOW PATTERNS IN NATURAL CIRCULATION ARE DIRECTLY RESPONSIVE TO RESTRICTIONS, AND MAY WELL SHOW LOCAL TWO-PHASE TRANSITIONS AND RELATIVELY HIGH VELOCITIES AT PARTIALLY-BLOCKED HOTTER REGIONS.REDISTRIBUTION OF CORE DEBRIS IS ANTICIPATED W4HEN THE PUMPED FLOW IS TERMINATED. CHANGES IN THERMOCOUPLE INCORE TEMPERATURE DISTRIBUTIONS SHOULD BE EXPECTED.NOT ALL THERMOCOUPLE READINGS CAN BE EXPLAINED BY HYDRAULIC PHENOMENA. SOME THERMOCOUPLES, PARTICULARLY NEAR THE CENTER, ARE CURRENTLY INDICATING LOCALIZED HEATING EFFECTS AND ARE NOT MEASURING BULK FLUID TEMPERATURES.. SINCE THE PRIMARY CONCERNS DURING THE TRANSITION FROM ' PUMPED FLOWS TO NATURAL CIRCULATION ARE ADEQUATE COVERAGE OF THE CORE AND BULK COOLANT TEMPERATURES BELOW SATURATION TEMPERATURE, IT IS REQUIRED THAT.AT LEAST 10 (TEN) INCORE THERMOCOUPLES HAVE READINGS BELOW SATURATION TEMPERATURE FOR THE SYSTEM PRESSURE (FIGURE 5). -ADDITIONALLY, NO TWO IMCORE THERMOCOUPLES SHOULD EXCEED 800 0 F. : ANTICIPATED CORE TRANSIENTS ARE VERY SLOW. FIGURE 4 SHOWS THAT IT'WILL REQUIRE AT LEAST 45 MINUTES TO ONE HOUR TO RAISE THE WATER TEMPERATURE IN THE CORE' FROM 200 F TO SATURATION TEMPERATURE FOR DATES BETWEEN 4/10/79 AID 4/17/79. ADDITIONALLY, THE CORE ADIABATIC HEATUP FROMl 200 F TO 1000 F EXCEEDS ONE HOUR FOR DATES' AFTER 4/12/79.HENCE, NO CORE COOLING PROBLEMS EXIST FOR AT LEAST THE FIRST HOUR OF TRANSITION TO NATURAL CIRCULATION. THE "HOT-LEG'" THER1lOCOUPLE SHOULD SHOW A TEMPERATURE INCREASE AS NATURAL CIRCULATION IS ESTABLISHED. TEMPERATURE INCREASES IN THE T. RESISTANCE TEMPERATURE DETECTOR (RTD) SHOULD BE OBSERVED WITHIN THE BOUNDS OF-THE TIME TO-,SATURATE THE CORE AS SHOWN IN FIGURE 4 (I.E., 45 MINUTES ON.4/10/79; 2.5-HOURS ON 6/6/79).-.w ~~.-.$ : 5:40 P.M. APRIL 10, 1979.G. A., MEYER ADDITIONALLY, THE MAXIMUM TEMPERATURE OF THE HOT LEG RTD SHOULD NOT EXCEED 250 F FOR PRESSURES ABOVE 500 PSIA AND 180 0 F FOR PRESSURES NEAR AT.MOSPHERIC. THE CORE AT SHOULD BE LIMITED TO 150 F.

SUMMARY

REQUIREMENTS FOR TRANSITION TO NATURAL CIRCULATION:

1) INCORE THERMOCOUPLES

-- AT LEAST 10 (TEN) THERMOCOUPLES MUST READ BELOW THE SATURATION TEMPERATURE CORRESPONDING TO SYSTEM PRESSURE.-(FIGUPE 5).-NO TWO INCORE THERMOCOUPLES MAY EXCEED 800 F.2) HOT LEqRTD'S -- THE HOT LEG RTD'S MUST INDICATE A TEMPERATURE RISE WITHIN THE TIME REQUIRED TO SATURATE THE CORE' (FIGURE 4) AND THE MAXIMUM TEMPERATURE--- SHOULD NOT EXCEED 250 F FOR SYSTEM PRESSURES ABOVE 500 PSIA A-ND 180 F FOR PRESSURES NEAR ATMOSPHERIC. THE CORE AT SHOULD BE LIMITED TO 150F.LLL:JFB :nw cc: J. S. TULENKO FUEL ENG. UNIT MANAGERS CORE HOT SPOT TASK FORCE ATTACHMENTS (5 FIGURES) 0 0 L4t 10 X 10 TO 54 INCH 7 X 10 INCHES.WEh KEUFFEL & ESSER Co. MADE IN U.&A.,,I 0 46 1320 4 ii I.... .... ... 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H i ll.: II' *j!I I: II .4 I ,jI., ... .. .1. ... a4 .1 ,,...... .. .4..44 ... 44. ...22 ' i sI.i ,: :t iiii1 I '2:2.::: ~ 2::1 '::/O9 I::; ~4i. jIi it:: I::;2::1-::::::l .._ ._. _ _ ..:::: :i:: :::- !: ::: :.!:::: : :1: : : ::::: I: .. E .fI 1i_ _ _ _ _ _ _ _ _ _ e: L i !; : is 2 : 2 .: 2 : 1: 1 i 1:::: 1.!:: ];::! 1!::: 1;::: 1:;:: : -;: 1:;:: 1::! .. I:: , :; I: ; :!:; :!.,! :% , :::: 1, , " .. .... .......;J Jo. ., ,'-'. .. .,:::: I': i: I..-: 1:::.. 1: i ill::!: 1: -I I::: 1: I ::::: :-.: -::: :I-: : .. !!!! !HHAN, :".,: 1; -.; .. ... ....... ...N .:: I; ..:.: :::: :1:: :.;. .1" .:j.. !, .. .... .:.. ..,:.I!! !,I I "I" T"T .1 '. .....I ..._ ._ ....... -... ..-Is THE BABCOCK & WILCOX COMPANY .cc: J. R. Gloi G. A. Mey4 POWER GENERATION GROUP G. A. Mam P. E. Mame idemans ar:)la ro I E. OELKERS, SYSTEM! RELIABILITY & CONTROLS UNIT From Pfuo P -I f:, r TM-J. A. Weimer H. D. Warren C. T. Rombough R. V. Debars 9DS 663.5 T. L. WILSON, SYSTEM RELIABILITY & CONTROLS UNIT Cust. File No.TMI-2 -o r Ref.Subj. RELIABILITY AND UNCERTAINTY OF THERMOCOUPLES FOLLOWING APRIL 5 1979 LOSS OF FEEDWATER TRANSIENT sii tstto d (owv *an customnf srd e"* tvbjtsI only.Acknowledgement: 'I have gathered the following information through conversations with several individuals having experience with thermocouples that have been subjected to extremely high temperatures. I am indebted to P. E. N'Jamola, R. H. Stoudt, v.and Tom Kollie of ORNL for their help and cooperation. Description of the Problem: The incore thermocouple design at the TMI-2 core is a grounded junction Chiomel-Alumel, detector. The wire diameters are approximately 10 mils, the sheath is 62 mils OD Inconel tubing, and the insulation material is A1 2 0 3.I have gathered information on reliability and uncertainty of this type of thermocouple after being subjected to extremely high temperatures (> 2000F).The problem of gross failure (open circuit, sheath failure, new junction or other failure) would be indicated by no reading or extremely low, erratic reading. No detectors are giving indication of gross failure. Given the survival at present conditions, the prospect for continued operation is excellent. The primary questions are the following:

1. Are the readings accurate?2. Are the errors in the readings consistent with the hypothesized scenario of the transient?

Decalibration Phenomenon: A phenomenon observed by Dr. Kollie in thermocouples having experienced extremely high temperatures is a deteriorated state in which the thermocouple gives a stable but inaccurate reading. The effect is called decalibration. The decalibration error is random but does follow some trends. The primary dependencies are the following:

1. Sheath diameter The larger the detector, the less susceptible to decalibration.

2. Sheath material Inconel is better than stainless.

The decalibration E. OELKERS 2 APRIL 5, 1979 error for Inconel is usually negative. Measured temperature is lower than actual.3. Temperature. Decalibration error increases with temperature. (10CC roughly doubles the reaction rate of the mechanism.)

4. Time of exposure Decalibration error is roughly linear with exposure time.5. Temperature profile Decalibration occurs along the length of the thermo-couple. Hence, the error depends on the profile of* the elevated temperature and on the profile of a subsequent measurement with the thermocouple.

Other dependencies which are not known specifically but are expected to be small are insulation material, length of detector, and wire diameter.Experience with smaller diameter detectors (20 mils OD) but similar in other characteristics (grounded junction Chromel/Alumel thermocouple with Inconel sheath) show essentially zero probability of survival at 1100'C for 10 to 100 hrs. without a measurable decalibration. The decalibration error tends to be negative (reading lower than actual) for this test; however, one sample gave va positive error. The magnitude for worst case error (remembered) was -50 0 C.larger decalibration errors would be expected for higher temperatures; The primary mechanism for decalibration is migration of constituents primarily chromium in the detector through vapor phase transfer. The chromel lead loses chromium and the alumel lead gains chromium. Belfab tests indicating little decalibration for 4 hour period at 7000 0 F are not necessarily conclusive since the temperature gradient of the high temperature state does not simulate the transient environment nor did the test condition simulate the present environ-ment.Conclusions: If it can be determined that the thermocouple is reading accurately now (for example, if it agrees-with outlet RTD's), this implies that its readings are believable throughout the transient. If thre is reason .to believe a thermocouple is decalibrated, there is a high probability that actual temperature is greater than its indication. Hence, the cluster of high readings should not be disregarded. Since a high percentage of detectors survived the transient, *the maximum tempera-ture did not attain a level for widespread failure.Recommendations:

1. Perform testing simulating transient conditions to determine the temperature threshold for gross failure in TMI-2 design detectors.

2. Perform testing to quantify magnitude and direction of decalibration errors for a range of temperatures simulating reactor environment.

TLW:ae THE BABCOCK & WILCOX COIMPANY*POW'ER GENERATION GROUP. .To lw.Engineering Operat-i-ons- -Manager irom-J. T. -Wilise ~ -7i~THE-79-179-ODS 66-.5 Cust. \7_ -File No..or Ref.Subj .Date Action Item 143. April 6, 179-~~ ~ ~ ..pi ,17l aa l.ett .r to 'cver o"* cstom*t end oe* subject only.

Reference:

C. T. Rombough to Engineering Operations Manager,.Action Item 143, 4/5/79.The purpose of this memo is to elaborate on conclusions no. 3 and 4 in the referenced memo.. The increase in selected thermocouple readings is no cause for concern. The temperature is still 200 F below saturation temperature. The'increased readings are caused by two factors. The first and smallest effect is a 4° increase in the core inlet temperature. The primary cause for the chpnge in thermocouple readings is the change in the flow distribution caused by the shifting debris in the core. This' was vividly.demonstrated when the Al pump tripped. and -the .A2. pump..w-as started. --I would anticipate that some thermocouples would continue to change for several days until the debris redistributes into a stable configuration reflecting the change of coolant pumps.0. C, ... .Conclusion number 4 is inaccurate since resistance readings wiill show wide variations from thermocouple to thermocouple and also the-:readings will depend on, whether the chromel or alumel wire resistance is being measured. However, if needed, we can state.the following: A test can be performed to determine whether a 4ross failure of a thermocouple has' occurred. Fo'r a good thiermnocouple the resistance between one T/C lead and ground should be approximately 750 ohms while the other lead to ground should measure approximately 300 ohms to ground.JTIV:mp cc: F. E. Unit Managors___ J. S. Tulenko Shift Technical k ader THE BABCOCK & WILCOX COMPANY POWER GEMERATION GROUP-To 1, Engineering Operations Manager rom_ _ T .¢_-,I_ r l5st -nnt EDS 663-5_.] , Kooun ru+/-z Cust. File tNo.GPU or Ref. ONCP-79-038 Subj. Date 3:30 p.m.B&W's View of Increasing Incore Thermocouple Readings (Instruction 143) April 5, 1979 I 1b6W loiter to -cor on. cszteort and Ofte sabjecI enly.As requested in Action Item 143, the increasing thermocouple data has been reviewed by both NFGD and LRC personnel. These personnel have included J.B. Andrews, G.A. Meyer, J.A. Weimer, J.T. Mayer, T.L. Wilson, E.T. Chulick, P.E. M~amola, R.A. Copeland, J.W. Ewing, H.D. Warren, and J.G. Brown.A summary of the pertinent data and their conclusions is attached for D.H. Roy's response to Bill Lowe.CTR~dlw I*.' .*3-.. 6 q-v a I... .I Incore thermocouple data from Tta-2 have been evaluated. As shown below, six of the thirty thermocouples for which we have data have shown a temperature rise of'7 -33 0 F over the past 5 days (117 hours) or 1.4 -6.8°F'per day." TIC at T/C at Net Increase Location' 0845,3/31 0542,4/5 Increase (OF) Per Day (OF)13C 13F 13H 11G'12F 1lL 290 298.310' 427' 303 302 297 307 320 445 327 335.7 9.10 18 24 33 I1.4 1.8 2.1 3.7 4.9 6.8.Temperature vs time for locations 13F, 12F, and 11L are shown in the attached figure.The following conclusions have been reached based on this data.1-. 'There is no mechanism which has been postulated that would cause docalibration to the e tent shotun by. the data. Therefore, it is concluded that the thermocouple readings are accurate and that true' temperatures are being monitored to +/-50.2. There is nothing in the fuel, structural materials, or fission products which would cause a chemical reaction that would result in the observed rate of temperature rise.3. As indicated in the attached map of temperature changes for'30 selected locations, the increasing readings are located preferentially in the core-'(in the center of the right half). This leads to the most probable explanation that a very gradual flow redistribution is occurring; either increased flow blockage in this region or decreased flow blockage in another region causing temperature increases in this region.4. If M4et Ed questions the thermocouple readings, a simple test can be performed. The resistance between the T/C lead and ground should be approximately 750 ohms. CAT.I,. i ., .*i ....~~. ... ... .. ..... ..... .........I LI. :.... .~ ,...... .... ... ...... .. *......A .1... .. -..*I. ........... ... * .I I I * ~

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..... ...... ....- t Il. .. ........____ I I 2%% //L. I, l iii .W ad dO s o;4 10 0411 --o4oI- ., ij lg.I eM M%.%KV , -* -:- 1 I*2A 0 , ch4AMSE ZN Ir/6 R 2 3 4 5 6'rot t D//V -ap odr33 No 0A/7 8 . 10 11 12 it .1 14 15'/5.i'4"; .e \.~ ..--I , -9. -B-a --I 4. -4 .1 I-.4 I D-l~~tlI'I- --I-47 so F G H.K I_,.._ _ -. _ .T_._1 L -he -ie Js~i-t 0 -..30.zI-I-.---_ _ -._ _ __-- -,, ._+4-'-3)i i i i ..i --I..... ----. -. -M I-20.3 2A3 N--20 .3 0 1,.I..I*- , , .-- _ ------ A_p I I 0 I .....II-i_ _ .,IHE BABCOCK & WILCOX COMPANY POWER GEEfRAT ION GROUP To A.MEYER, MANAGER, T-H ENGINEERING UNIT THE-79-171 J. A. WEIMER, T-H ENGINEERING UNIT, EXT. 3236 IDS 663.5 Cust. .File Ho.or Ref.Subj. Date RESPONSE TO THERMOCOUPLE REQUEST APRIL 5, 1979 This letter to cover one customer and one *ubject oelt.Past experience with thermocouples show two.:basic things. First, they track fairly steady when core condition are not changing (ie. 100% steady state conditions). Secondly, they respond to local changes relatively accurate. A one to two degree change in thermocouple reading has occurred many times without any obvious core change (or local change) conditions. Greater deltas than 2 0 F usually indicate another measurable local or core change.The .two mechanisms for thermocouple changes are obviously power distribution or flow distribution. A few random local changes are.usually due to local power changes. A large area group change indicate flow redistribution. > Assuming local flow blockages (in the center of the core a decrease in thermocouple reading would occur if and when the blockage was decreased and if the "hot" conditions-were due to local heat sources (ie. fuel pellets) the temperature would decrease as the source burned out. Furthermore, as blockage decreases in one region of the core the outlet;.flow distribution wouldtend to flatten causing temperature increases in other-portions of the core.JAW/sgh cc: J. S. Tulenko FE Unit Managers J. F. Burrow jW , ~ STrzif3unr3 .I (& 74 3 SRoM C.r. 3-tioorH .( 3 ?0)-3 ..I *1 -- --FiLt NO. OR REF.UZ EAi rDS / AM 2' ;

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c__ ,T tt __Cr4 _jL>taJD .S 6;-' a AHyd-_P e~rev. __ _ _ _ _ _ v ~_ _ -r. 'V-TTv- .rrTT Ti-rT .-.* ~ GE14ERATIONl GROUP To 'W;2C-J!lRYOa.R < * .... .-D. H. ROY --' ' ' 'DS 663.Cust. ..File.lo.or Ref.Subj. REQUEST FOR B&W'S VIE'.4 OF INCREASING ..Date INCORE THERMOCOUPLE READINGS APRIL 5, 1979 t his lotlerms, a .s r e o. s"tomer U"d' iee sbject i inly.NOTES TO ANSWER TO BILL LOWE'S REOUEST TO D. H. ROY AT 0735 ON APRIL 4 REGARDIING B&W'S IDEAS AhND OUR CONCERNS REGPRADING SO;!E THERMOCOUPLES SLOWLY IINCREASHIG III TEMPERATURE , IL~rrw Iu¶. ...." The responses provided is inadequate for transmission to GPU. We should take the following steps: , (1). Report the increases 'as we see them in the data -assembly location,-temperature change per unit time over some time interval.(2) Provide the uncertainty value which vie would assign for these loca-tions...(3) Have our radiochem'ists get in' touch with expert fuel material people to determine if there are any break away chemical reactions or any otherexplanation which might be provided to explain the temperature behavior which is observed.' It.is not necessary to state that vie wi1 continue to monitor thermocouple changes and report. them promptly as GPU is also tracking.them routinely. D. H. Roy DHR:dmd -II-clecon from-Bill Lovc to D. R. Roy April 4, 1979 t 0735'Tclecon from.Bill Lowe at 0735 informed us that over the past 36 hours* .*.some trace in thermocouple readings have been observed, it varies from assembly to assembly with some going up by as much as 9 F. The absolute temperature still is o.k. but they do not like the trend. It could be associated with changes in coolant temperature and or pressurizer level.Me requested that we consider various possibilities for thin h-baayor im.U.a break away chemical reaction. We told him that our data for thermocouple readings at the 05Z3 hour measurement was just coming in, we will take action to determine whether there is a chemical reaction which can account for this and tr 'to corelate with changes in the state of the reactor.Sin ....D E o; .., .I indD l.Ry .*. ......*.. ., *. ...* .a 4**.* I -.-.* .I FROM .I................ ..1 ---.. ..b > 9 w--z FlL NO OR REFI I rIL: NU. UK ALP: SU51 DATE~uaj._______________________________ 4 __________ 09o/2 __o_4s.(LJ ____lD____l4D(S < o SW > 77EC C)l 3 .__ C ____C 02oor/7, orbo 4-1 ___T ~ c _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___ji. .iq .-__________ _~l _ _.i....._ _ _ _ _ _ _ _ _ _................... _ _ _ _ _ _ _ _ _ _ _ _ _ _' .> __ _ _ _ 33 _________4__ _13C .7 0--F. /... O _ ._ .A r _ __ _ _ ___ A t__O.tD. eY?. ._s__° ' o°, <AA_,_ i.s .. / ._-os- _ __ F ...... .. .;.et- :.'.-:1 '-ie.- T t : 1_-_-- ' _ _ _ _ _ _ _ _ _' L_ ._ ._ _ ............... .,.... ..__ _;._-__ _ _ ___-s _l-

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TMH YJAE3COCK Cz WILCOX CO._, ._ .CL ..0.,,C&,t--l- -r/c-* d X ' )9 oo, -/s o4-D1, 0 'i-A-5 SPND STRING NUMBERS AND J LOCATIONS-, 177 FA CORE I STRING NO.-r__ __ .; O- .--V~31 30-..+A 30-, ---*3?.w1li 457 4q-~ -.* -,- --' -- -r-- 4. -r -1 .___Z'5Iz Si2 34 I'-47, 5ri 26 I. ----. -> -.-.-C.> -8 .4 t 4 -1t I I. I 4I10 iTT.+35 g.I l -I+3

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I_______---mt ,r1 77ctv w x iLr. 7*S., tpR -r ;J -.SPND:.STRING -NUtMBE RS -AND LOCATIONS -177 FA CORE STTRING NO.I..C --I f tvoe (al, 3Z L1.4 ----: -*f -* 31 30 31 Zi I 30 ,/--.-'-I --. ------Z8-. .-1 --I -------L 1l 34 q'3,04..._I//---~ ~ + --f -4 i ~ ~6Z7 q V4_ 3 1 M ..Z2_wt__3-22 313 _ 3z97 ___r3 1.._> ..10 .71~ 313221 -101 1_Z1 L. ._ ._ Z ..r r1-l 37aoI !C~--°I I a SZS371-l %5q_ _.__12& 5-5-l 1111 41 - -T --.-1~~ -L~l. I, 2 3 '/ S 7 8 71 10 U 12. 13 if z EZR I Jon o0.A r 'Evaluate fuel pin contact with Incorc thcenril couples The evaluation was performed assuming all the pellets from one grid span are , evenly distributed across the upstream grid. Avcrage temperature of the UO2 mass is -1500 0 F. Enough heat is generated to produce boiling in the annulus bctween the instrument string and the guide tube and possibly some superheat. Heat would be transferred to the instrument string which could be transferred axially by conduction to the thermal couple. This would produce a T-C reading higher than the surrounding coolant. This indicates that it is possible that a conglomerate oG pellets with the proper size and location could produce T-C readings in the range of temperatures which are being recorded (100-600 0 F above coolant temperature). C. D. .organ M. Montgomery C. It\.G. A. Mleyer 0

t. -.- -IfHE BABCOCK & WILCOX COMPANY POESEM GENERATION GROUP ?f --IM /AIA-P. tER F To I C. C. ENGLAND-rom A. B. JACKSON ZDS 66 Cust. .' Fi Ic No.TMI -2 or Ref.Sub3. Date 12:40 A.M.REQUIRED FLOW4 FOR CORE COOLING APRIL 10, 1979l li letter, to tover one customer ond one swbji' only.I

REFERENCES:

(1) MEMO, C. C. ENGLAND TO R. B. DAVIS, "ALTERNATE DECAY HEAT-SYSTEM4 REQUIREMlENTS," 7:22-A.M., APRIL 9, 1979.(2) MEMO, G. A. MEYER TO J; D. CARLTON, "MiNIMUM CORE FLO', -LONG TERM COOLING," 1200, APRIL 4, 1979 (3) MEMO, A. B. JACKSON TO C. C. ENGLAND, "CORE FLOW R EQURED FROM HPI SYSTEM," APRIL 8, 1979 AS REQUESTED IN REFERENCE (l), THE UNCERTAIN;ITIES INCLUDED IN THE CORE FLOW; REQ-kQUIRE-MENTS BY T-H HAVE BEEN RE-EVALUATED BASED ON UNCERTAINTIES PROVIDED BY CEINTRA\L ANALYSIS. UNCERTAINTY DUE TO CORE BYPASS HAS BEEN REM1OVED AND THE CORE FLOW REQUIP MENTS FROM REFERENCES (2) AND (3) ARE REPLOTTED ON FIGURES 1 AND 2. THE CONTROL ANALYSIS-UNCERTAINTY OF 5.8 TIMES THE CORE FLOW;I REQUIRED MUST BE APPLIED TO THESE CURVES.A. B. JACKSON ABJ:dmd L CC: JS TULENKO FEUM I I I I no;, .. ...-.. ---EUFFEL & ESS, "W, , ,,.1*-I I Ito 1cv hl 11 '":I.I .i~ .:1. 1 ..... .j...~I I ..j....I. ... ..... ...iiI d il~ .... I... ...1.___...___..._ ........ .1.I. .....to'II I I I ..

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.... .....J.... .. .............. .....-- -- -- -- --ILI.. ... .... *: K I I I. .... .. .K IV... ....K. ... ........ .K. ........ .... .... .... ..... .. ... T E BABCOCK & WILCOX COMPANY.4OWER GENERATION GROUP x4 INSTRUCTION 11310 LICS To -B...R. B. DAVIS, CONTROL ANALYSIS, A. L^.,JACKSON,,THER,1A7\HYDR.At From C. C. ENGLAND, LONG TERM COOLING&DS 663.-Cust. File flo.or Ref.Subj. Date TIME: 7:22 P.m.ALTERNATE DECAY HEAT SYSTEM REQUIREMIENTS APRIL 9, 1979 lj 1hlitis 1 to coSve one customer ond one subjecl only.R. B. DAVIS' MEMO OF APRIL 8, 1979 RECOENDS THAT DECAY HEAT SYSTEM FLOW REQUIREMENTS BE SET AT 5.8 TIXES THE CORE FLOW REQUIREMENT. I UNDERSTAND THAT THE ATTACHED CURVE FROII THERMALNL HYDRAULICS ALREADY CONTAI'I'S. SOME CONSERVATISM (ON THE ORDER OF A FACTOR OF 3) TO ACCOUN*ft FOR UNKNOIN'S IN THE CORE.I REQUEST THAT YOU REVIEW YOUR CALCULATIONS WITH THERMAL HYDPAULICS TO ASSTRE THAT TIE'.:tICONSERVATISMSS" AREN'T 13EING ADDED TWICE.A D.. .FLOW RATE MUTUALLY AGREED UP ON BETWEEN CO'NTROL ANALYSIS AND T.H.E.SHOULD BE ESTABLISHED. PLEASE PROVIDE THIS BY 0800 ON 4/9/79.CCE/CMW'ATTACILMENT -, C.SCOCK & W4ILCOX CMIIANY /.AT I C UP I. _ *\.** ~ ' ~ ~ *CC ENGLAND * \f rom Ar i U UJ .REQUIREMENTS FOR ALTERNATE DECAY HEAT SYSTEM I 1his 1elgefe to g.r en.' custamot *ad one sbject only.I

REFERENCE:

CC ENGLAND, RB DAVIS, SAME SUBJECT, APRIL 5, 1979 IN THE REFERENCE, I RECOMMENDED THE DECAY HEAT SYSTEM FLOW REQUIREMEE-r BE 3 TIMES THE CORE DECAY HEAT FLOW REQUIREMENT. THIS MEMO REVISES THIS:.REQUIREMENT TO 5.8 TIMES THE CORE FLOW REQUIREMENT. IN MY LATEST ANALYSIS, I IMPROVED ON MY CALCULATION TO INCLUDE (1) A BETTER HYDRAULICS MODEL, (2). A HIGHER CORE RESISTANE .(83 x 108 psV Ibs/sec)2)*POVIDED BY PS BARTELLS, AND (3) A CORE BYPASS FLOW OF 27%. THIS NFWd DATA RAISED.THE REQUIREMENT TO 5.8 TIMES THE CORE FLOiW REQUIREMENT. ATTACHED IS GEORGE MEYER'S CORE FLOW REQUIREMENT ON WHICH I ADDED A NOTE ON THE DECAY HEAT SYSTEM FLOW REQUIREMENT. THE DELTA P (FROM THE CORE FLOOD NOZZLE ACROSS THE REACTOR VESSEL AND UP TO THE DECAY HEAT DROP LINE) IS STILL NEGLIGIBLE (APPROXIMATELY .02 PSIA);js..B __ _ 7*- --J...e z , THE BABCOCK & WILCt)X COmPANY POWEr GEtERATION4 GROUP 1*cc: D. A. Farnswoith R. 11. Stoudt RECEIVED'liPRi 0 1979 rt- A 6DS 663.5-G. A. MEYER._FUEL ENGINEERING From r J. R. GLOUDEMAIIS, THERMODYNAMICS UNIT, TECHNICAL STAFF.: M* YFR Cust. File No.* or Ref. 845-7952-01 Subj. Date COOLDOWN PRESSURE APRIL 10, 1979 I h~i letter tocover one cwtomr end _c. mvbject only.I COOLDOWTIN PRESSURES FRON ATMOSPHERIC PRESSURE TO "v 1800 PSIA HAV\E BEEN CONSIDERED. THERE ARE MANY SYSTEM RESPONSES RELATED TO PRESSURE, THE ADVANTAGES OF.COOLING AT HIGHER (UP TO ' 1800 PSIA) OR LOWER (DOWN TO 'v 15 PSIA) PRESSURES ARsE:,.1. ADVANTAGES OF COOLING AT HIGHER PRESSURE a. GREATER MARGIN TO SATUPRTION TEMPERATURE.

b. LESS STRAIN ON (OPERATING)

RCP's.c. BECAUSE OF DECREASED EXPANSIONl OF FLUID DURING VAPORIZATION (SEE ATTACHED FIGURE), DECREASED FUEL OR FUEL-DEBRIS DI-SLOCATION DUE TO THE MEATING ANID VAPORIZATION OF TRUPPED WATER.2. ADVANTAGES OF COOLING AT LOWER P RESSURE a. BECAUSE NATURAL CIRCULATION IN RESTRICTIONS AND/OR AT "'HOT SPOTS" MAY REQUIRE VAPORIZATION TO ACHIEVE THE NECESSARY STEADY-STATE DRIVIING HEAD, AND VAPORIZATION EXPANSION INCREASES WITH DECREASING PRESSURE (SEE ATTACHED FIGURE), LOWER PRESSURE INFERS O10RE ADEQUATE COOLING OF THE MIORE-RESTRICTED FLOW REGIONS IN THE CORE.b. POSTULATING NATURAL CIRCULATION AND VOIDING AT CORE-FLOW RESTRICTIONS AS IN"2.aa", AND NOTING THE DECREASE liN SATURATION T1S ERATURE WIT11 PRESSURE, G .A. MEYER 2 -,PRIL 10, 1979 THE (FUEL) DEBRIS OR FUEL IN THE STARVED REGION iS AFFDISRDED A LOWtER SURFACE TEMPERATURE (WITH LOWER PRESSURE).

c. LESS PRESSURE-INDUCED STRESS ON; BOUNDARY COMPONENTS.

d. LESS BLOWDOWN (TO ATMOSPHERIC)

LIKELIHOOD AND SIRESS.c. FACILITATED SHIFT TO THE LOWER-PRESSURE DECAY-HEAT-RPOVAL SYSTEM. *I AM CERTAIN THAT LONG-TERM COOLING SHOULD NOT BE A.CCO!MPLISIIED AT THE HIGHER PRESSURES CONTDXPLATED, 1000 to 1800 PSIA. THE OINLY SIGNIFICANT AD'VANTAGE (la), IS OUT-1WEIGIIED BY ITS COUNTERPART (2b); i. e., RAISING PRESSUIRE RAESES T AND DIRECTLY RAISES THE SURFACE TElPERATURE OP: THOSE FLOW-RESTRICTED SAT.)DECAY-HEAT REGIONS WHICH REQUIRE BOILING FOR THEIR HEAT Tit-.NSFER MEC1'-ANISM. IWIILE RCP' s ARE OPERATING, I RECOMNEND THE LO.,TER B.fAX: O'Y THE CURlRENT PRESSURE RANGE, i.e., ' 550 + 50 PSIA. PRESSURE SWINGS SUOULD BE AVOIDED.(COOLDOIW SHOULD 'BE VERY GRADUAL). WHEN "COOLED DOWNt" (TO APPROXIMATELY 100-200 F), AND WHEN RCP's ARE NO LONGER AVAILABLE OR AE NOT DESIRED FOR BACKUP, SYSTEMI PRESSURE SHOULD BE SLOWLY REDUCED TO. APPROXIMATELY 100 PSIA (T SAT" 328 F). THIS LOW.ER PRESSURE INCREASES THE DRIVING HEAT AT VOIDING LOCATIONS AND DECREASES THE SURFACE TEtIPE}'ATURE OF BOILIN-3-LIITED FUEL, WHILE MAINTAINING MORE THAN 100F "MARGIN" TO SATURATION..R. Gloudcmans VERIFIED 4/10/79 by 1D. A. Farnsworth /,,w.,,.,. /,,/dh,,, -~ , So A~"'i r .-' I R ( p ) C', t, % "', I ,- o. , P)i.' '0 \t I ' .i I-J.o C '.* * .. *. *1 ' ' I. 1 ,* ~.-* ., .i. I. .ii i i ) ;i 'l ; '. .ii *q I o --........... _v.z --j! .I ..5-.S~~~~ .I :-.!...1'g ' ..I .;t .- I,- -, a I f-, ~ &.~c .~. -I I .t , S ...j _ 1._ __ i _** 1_* 1 I' l '.,t1 .L-.i,- i j, ..I--.-I* I+/- KI 1T7 i--i.1 ITI I .-.l*I..I;LiI 2* *-1.-1.-I I , -.i.-* .r.. 5 I I! I:.I I I t I.1 I I I S I-J I

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WILCOX COMPANY x POWER GENERATION GROUP .. .. X THE-79-193 To -G. A. MEYER, MANAGER, THERMAL HYDRAULIC ENGINEERING UNIT From J.4H. g&NES,(THE UNIT, EXT. 3239 SD$ 66: Cust. .File o.¶11112or Ref.TMI2 c.Subj. Date ADIABATIC HEATUP RATES APRIL 10, 1979.this letter to cower on* ,sgtomer and on* subject enly.FIGURE 1 SHOWS THE ADIABATIC HEATUP RATE AS A FUNCTION OF TIME IN DAYS AFTER THE LOSS OF FEEDWATER EVENT. THIS CURVE ASSU.:ES INSTANTANEOUS CORE UNCOVERAGE (NO HEAT REMOVAL FROM THE FUEL). THE CALCULATION WAS PERFORMED FOR A FUEL ROD ASSUMED TO BE INTACT.FIGURE 2 SHOWS THE TIME IN HOURS TO HEAT THE FUEL FROM 200 F TO 1000 F USING THE HEATUP RATES FROM FIGURE 1.FIGURE 3 SHOWS THE DECAY HEAT (CORE POWER) AS A FUNCTION OF TIME (SEE REFERENCE).

REFERENCE:

MEMO, J. R. BURRIS TO J. D. CARLTON, "DECAY HEAT CURVE," APRIL, 2, 1979, NSS-6.JHJ/CDB CC: J. S. TULENKO F.E. UNIT MANAGERS CORE HOT SPOT/FLOW BLOC1lAGE TASK FORCE Q/A Both method and ;calculations have been independently checked and found to be correct.77 t (.0'.A!2 t4 1.0.... ... .. ..... .. ........... ...... -7 I.... ....... I I I ..______L t ~ I........_...__....-h17 V 4- ~~..7 IH---- ----_ _ _ _ __._ _ _ I.j IC, oi~.h a..ox r~C..W-------- ----- -----I --=- -.-I '=j..__ _ _. ....__ _ _ _ _ _. ....._ _ _ _ _ _ _ _ _ _ _.... ...... .......L.FE.. .........._ _.. ..o. %-in- il ...._ _ __..... .._.............. _.. .._ _ __=_ _--_ : : --".. .a./ .......Ar/ __... .... .... .... ..~ ... .. .. ... .. ... ... .. ... ... ...7II.j ...I ....j..... ./. .. .... .. ....I.. .. ...._ _ _ _ __.. .._ _ I -... .. --;'7i -7. 0 LO~-.4 (-9 I'J Z 6.U.tf iow ow~-r C'* ....... ... ...7I 7... .... ... ......J -~~.-... .... -.. .... -- ...] I i ..i l ......_Ii.. .... .... ..I ) A A ~ /... ......... ...P .... .r... ........ ..._.................lr r j. .......... t 2..--- -I v nv W --f j 4 t 9 % J9 I j ) % .1... ... ..... .... .... .......J............w.I .I -': 1 ':.I.. ..I 9 l I H I I i: .I 1: i ;,; I-.-!; : i, .,.I.- ..;I I ;!.. :,!.I II:Hiiii 7 : .-I I ..ii!1ilk;

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.APRIL 10, 1979 Thi i.ettr to go.er sone CI1unem and one swbject- only.

REFERENCE:

CORE CONDITION TASK FORCE TO J.S. TULENKO, OF CORE CONDITION, 4/7/79 (1800)," 4/7/79"CURRENT ASSESSMENT -7:48 P.M.ONE POSSIBLE EXPLANATION OF THE INCREASED T.C. (LOCATED IN UEF'S) READIRGS IS IR ACCUMULATION OF U0 2 FRAGMENTS IN AND AROUND THE NIXINC CUP. THESE ELEVATED 0 TEMPERATURES, RANGING FROM "100 F TCr "'190 'F ABOVE THE COOLANT TE'PERATURE, ARE IN THE CENTRAL PORTION OF THE CORE. ACCORDING TO THE REFERENCE, THIS IS THE POSITION OF THE CORE ASSUtMED TO HAVE THE GRIEATEST D4A>GE.THIS POSSIBILITY WAS INVESTIGATED ASSUMING UO PARTICULATES WERE WITHIN THE 2 MIXING CUP AND AT RADII OF 1/2 IN, 1 IN, 2 IN, 3 IN, AND 4 IN. AXIAL CONDUCTION AND CON%`ECTION WERE NEGLECTED (GROSS FAILURE).THE RESULTS ARE SHOTWN IN THE ATTACHED FIGURES. FIGURE; 1 SHOWS THE TEMPER,%TURE. DIFFERENCE (AT (F°)) EXISTING BETNT.EN THE I.D. OF THE MIXING CUP AND THE O.D.OF THE INSTRUM1ENT STRING ASSUMIING ENTRAINED UO THE TEMPERATURE RISE THROUGH THE UO IS SM11ALL FOR ALL TIMES. FIGURE 2 SHOWIS THE AFFECT OF VARYING AX'OU";TS 2 OF FAILED FUEL OUTSIDE THE MIXING CUP. THE TEMPERATURE DIFFERENCE (LT (F°)) IS ,'FROM THE SURFACE TO'THE T.C. WELL SURFACE. (THE ONE INCH' WIDTH OF UO0 OUTSIDE 2 THE T. C. W4ELL IS SHOWN ON FIGURE 1 FOR CO'XPARATIVE PURPOSES.) -. 30 M. 5.G. 'APRIL 10, 1979 IT IS POSSIBLE FOR TIHE ELEVATED T.C. READINGS TO BE SOLELY DUE TO LARGE AGGLONERATES OF FUEL PARTICULATES SURROUNDING THE MIXING CUP. FURTHER CREDENCE TO THIS THEORY ARISES FROM THE CHANGE IN T.C. READINGS HIEN THE A-1 PUMP TRIPPED. THEREFORE, ANY DECISIONS UPON CHANGES IN CORE CONFIGURATION SHOULD NOT BE ELADE SOLELY ON THE BASIS OF I'ICORE THERMOCOUPLE READINGS.PJH:nw CC: J. S. TULENXO FUEL ENG. UNIT MAN.AGERS CORE HOT SPOT TASK FORCE ,. -4 II 77- 7.~I~~i. -. -i......4.... L.i7... ... ..- .. ...I_____ ............ 7 *- ;. .. .F .. ..it I7 I =i 7..4- -_ _77__ _ _ _ _ _ _ _ ~ _ _ _= 7____ _ .41. .. I ~ ~ /. ____ ____ ____ ____ _ =4_ _ _ _ _ _I I ...,-.-V.-.-7 -I-~ ..-7_ _ _L.L___ 4 .L.. ~ ..... .4........ .. .1-I I .I-.-----... .. ..... .. .. .... .. ... ... ... ... .. .... .. .._ _ _ _ _ ... .......____________ r- :::j:77777i7 7i-7. .7;:77 7..::7t-J W J a _ _ __I_ _ _s fi l -s r ..r'*:1 *I ...-. **I*' .**4*4*-I

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.~... .... ..* * * .~ ~4' I ~ : '............._ ....... I..... ..... ... ...1. .1: ..-.: ... ..... .;......;I. L .................... iA_ .3, A'iF BABCOCK & WILCOX COMPANY POWER GENERATION GROUP PFee-' I MI MAY1/ ,,C (ok0-C DISTRIBUTION rrOm J. A. WEIMER/R. L. HARNE SDS 6A S3 Cust. File No.TM 1-2 or Ref.Subj. Date BOILING CONDITIONS IN CORE APRIL 1, 1979 I Ihis: lettet to co.er Gs Cwstomert and one subject only.The following curve (and attached calculations) shows estimated time required to bring the water in the core to saturation temperature at 300, 600 and 1000 psia from its present 2800F energy level. This calculation assumes a no-flow condition with 161" of stagnant coolant available for heat transfer. A second calculation was made to determine additional time required to vaporize the coolant such that active fuel will be exposed to steam.JAW/RLH:cmw >Attachment -I la.---- ..j... ..... .:::t: .I... ... ... H___/ _ ... .. .._....... ..Cm .:r:-::: E1i 7 7-4 T:1zNP fl::= ~IL ~T Q~AI .-......... ......................

L---_ _ _ _ _ _ _ _4 JI-4-I 0-- -...................

1- .t pg-K --=T--- .--:= r-L =- -;Z-- ___ -"r-7 ..7.. ......... ... ... ............... 4...........iI H Iz......-... =777- I.. .......-I J.E..c.... .............. ............. W. ..-....- ...... ...... .......:;:8::\: ..... .........z ........z- -...... ..... ~~fI... ... ............ I. .. ........... .....VH-7.- -I L::.s .. ._.... .. .... .IA:.::.. ... :j.. .. ....I.. ... .. .. ............

: ~I ~ I......_ _..._ _.. .....ow ....... ....

Y ..-hjEI3AEXCOCK & WIL11COX COMPAN*Y'~ -1 *--". r, rt I r A Ir I f~kl f nrfl II D.L~irtK ul;-I~Xi-ck I I'.11" uiuur X J. D. CARLTON, PLANT DE-SIGN .TIIE-79-1.62 , Om G. A. MEYER, MANAGER, T-H ENGINEERING UNIT (3 2 18)505 663.5-File HIoto.CIs. -or -Ref .Subi.- Date Time 1200 M ININUM CORE FLOW -LO.NG TERM COOLINGI APRIL 4, 1979 I MI .11.,o to Cover one C.13ofloer end oen subjeci only.Re f rencce: 86-1100401-00, I"Minimum Core Flow -Long Term," April 2, 1979, L. L. Losh.The attached figure expands upon the information provided in the TCfCereiice to provide minimum long terip cooling flow -requiremecnts for. vario-Us proposed operating conditions. Each of the curves presentced accounts for coi'e blockage by the application of a saf-ety factor of 3.0 to the calculated flow requirement. The cturves provided are essentially the same as those determined this morning-by L. L. Losh arid J. R. Bohart, thus providing a QA verification of their analysis.GAM/ sgh ..cc: J' S. Tulenlko~- FE Uni Man ag ers C. ParksAz, Y- ..I,..V ........... .. ... ... .. ... ... .. .. .. ... -.. .. .... ..-.- ~ ................. ., .. ... -............... I .-........-.-... _Z .... .. ... ..... .. ..... .._ _ __7 .. .. .. .. .. .I I in 4 ih.R x---------------- ___---2 ....--..... .......---------....................2. ..b 2ut :. r.-4.... ............... THE BABCOCK & WILCOX COM'IPANY.*POWER GENERATION1 GROUP Toms 12 G.A. MEYER -MANAGER, THErMAL HYDRAULIC ENGINEERING From R.M. HIATT -THERMAL HYDRAULIC ENGINEERING ITE-79-195 SDS 663.5 Cust. File No.or Ref.Subj. Date CORE FLOW. DISTRIBUTION FOR ONE PUMP AND TWO PUIP OPERATION APRIL 10, 1979.lidsi letter to toeet en, ulto., end one cubjfet irly.ONE OF THE IMPORTAN'T CONSIDERATIONS IN ANALYZING THE TMI-2 CORE-BLOCKAGE IMPACT ON CORE COOLING IS THE FLOW DISTRIBUTION IN THE CORE.* A DETERMINATION OF CORE INLET FLOW DISTRIBUTION FOR ONE PUMP OPERATIONI WITHOUT BLOCKAGE WAS BASED ON- A REVIEW OF THE VESSEL MODEL FLO7 TEST (VIFT) DATA AND ENGINEERING JUDGEMEhT AS FOLLOWS: THE TRANSFER OF MALSS CAN BE MODELED SIM ILAR TO ELECTRIC CIRCUITRY. TiEl.FLOW PATHS CAN BE REPRESENTED BY A SYSTEM OF RESISTANCES AND THE FLOW WILL SELECT THE FLOW PATH IN SUCd A WAY TO EQUALIZE THE PRESSURE DROP ACROSS THE SYSTEM. THUS, FLOW HAS A "LOOK AHEAD" CAPABILITY THAT TENDS TO EQUALIZE THE POTENTIAL (AP) ACROSS A SYSTEM OF RESISTANCE. THE FLOW4 CILNNELS THEROUGH THE CORE CAN BE VIEWED AS A SYSTEM OF RESISTANCES. BASED ON THIS PRINCIPLE, ASSUMING THAT RATE OF CHANGE OF MOMENTU1S FROM COLD LEG INLET TO.HOT LEG OUTLET-IS THE SAME, AND AN INSPECTION OF THE VXFT INLET FLOW FACTORS CAN BE USED TO -IDENTIFY THE FLOW DISTRIBUTION AT THE CORE INLET FOR ONE. PUMP OPERATION. FIGURE 1 ILLUSTRATES THE CORE INLET FLOW FACTORS FOR 4 PUMP OPERATIONT FOR THE 177 FA PLANT. SUMMARIZED ON FIGURE 1 ARE THE AVERAGE FLOW FACTORS FOR EACH QUADRANT. TiE OUTLET PIPE IS LOCATED BETWEEN QUADRANTS Al AND A2 C. APRIL'10, 1979 AND 'BETWJEEN BI AND'B2. THUSs WITH BOTH LOOPS 'OPERATING' THE RESISTANCES ' .ACROSS THE CORE FOR EACI.QUADRANT WOULD BE EXPECTED TO BE ABOUT THE SME.FIGURE 1 ILLUSTRATES A SLIGHT BIAS TOWARDS QUADRANTS: Al A2, AND Bt.- THIS BIAS IS PROBABLY DUE TO FABRICATION TOLERAN4CES. FIGURE 2 ILLUSTRATES THE CORE. INLET FLOW FACTORS FOR TWO PlUMP OPERATION. NOTE THAT THE OPERATING PUMPS 4.AND BLARE ARRANGED IN OPPOSITE QUADRANTS. THE

SUMMARY

FLOW FACTOR FOR EACH QUADRLNT AGAIN ILLUSTRATES A RELATIVELY.. UNIFORM FLOW FOR THE QUADRANTS WITH A SLIGHT BIAS TOWARDS THE QUADRANTS CON -TAINING THE OPERATING PUMIPS. NOTE THAT THE OPERATING PJi.IP CO,1INATION IS A T10 LOOP OPERATION. THEREFORE, THE RESISTANCE ACROSS EACH QUADRANT SHOULD BE ABOUT THE SAME WITH THE RESISTANCES HIGHER FOR THOSE CHIAINIELS FARITHEST FROM THE PUMP MAD FARTHEST FROM THIE OUTLET PIPING. H1OWEVTER, THE .MANTUM OF THE FLOW DISCHARGED INTO THE CORE MAY BE SUFFICIENT TO OVERCOME LATERAL CORE RESISTANCES IN THE LOWER PLEN.'ll. THIS IS ILLUSTRATED IN FIGURE 3.FIGURE 3 SHOWS THE CORE INLET FLOW7 DISTRIBUTION FOR A TWO PUMIP OPERA&TION 4, AND B II THIS INSTANCE THE TWO PUMPS ARE NOT OPPOSING. NOTE THAT BOTH LOOPS ARE OPERATING. THUS, THE RESISTANTCE ACROSS EACH QUADRANT ARE ABOUT THE SAME. HOWEVER,' AS MENTIONED PREVIOUSLY, THE LATERAL MOMENTUM pF THE FLOW4 DISCHARGED INTO THE LOWER PLENUM FORCES FLOW TO THE ADJACENT QUADRANT WITH A2. AND 132 HIGHER IN INLET FLOW THAN 41- AND Zi- CONTAINING THE PUMPS.FIGURE 4 ILLUSTRATES THE CORE INLET FLOW DISTRIBUTION FOR A 1 LOOP 2 PUMIP OPERATION (Al AMD A2).' THE INLET FLOW FACTORS SHOW A BIAS WITH QUADRANTS Al 'AND A2 HIGHER IN FLOW. THIS SUGGESTS THAT THE RESISTANCE OF TllE:'CORE QUADIRATS FOR TIHE CLOSED LOOP IS IIIGII OVERCOMIING THE LATERAL MOMENTUM OF THE G.A. MEYER-3-APRIL 10, 1979 G.A. .EE .3 API 0. 1979. ..DISCHARGED FLUID AND RECEIVES LESS FLOW. FIGURE 4 SHOWS THAT ABOUT 4% MORE FLOW ENTERS LOOP A CORE QUADRANTS THAN ENTERS LOOP. B QUADRANTS. WHEN VIEWED WITH THE INFORRATION OF FIGURE 3, IT IS BELIEVED THAT THE QUADRANTS Bi AND B2 HAVE A HIGHER RESISTANCE OVER THE COINiLETE LENGTH OF THE CORE. THIS OCCURS DUE TO THE "LOOK AHEAD" CAPABILITY OF THE COOLANT WHICl SEES THE CLOSED LOOP AND THE LATERAL RESISTANCE BETWEEN QUADRANTS Bl AND B2 AND THE OUTLET PIPING OF LOOP A. THEREFORE, IT IS CONCLUDED THAT A COMBINATION OF FIGURE 3 AND.FIGURE 4 IS THE MOST REPRESENTATIVE OF ONE PUMP OPERATION.. FROM FIGURE 3, THE QUADRANT ADJACENT AND IN THE SAIE LOOP WILL BE BIASED ABOUT (1.5 -2.5%)HIGH IN CORE INLET FLOW. THE FLOW FACTORS SHOWN IN FIGURE 5 ARE RECOCMMENDED FOR ANALYZING ThI-2 ONE PUMP CORE INLET FLOW CONDITION. W/FFA CC: F.E. UNIT MGRS. l-X*J.S. TULENKO CORE'HOT SPOT TASK FORCE ..i 7-1-Iy -V~ RC~cs$..SPNqD STRING NUMBERS AND LOCATILQ-N-177 FA CORE.-I.__AN I.E At = 1.0/o3 , K.37.Po 1I43 CI 7 STMUN4 C, N AL.AX -{.0olt7--ii --,q6I3.q37 g6 7.V77.9-7 3 , Vo.96o-'C -- ¶ ----Zq 3 1o/ 7 1 0/0.q37.q77.q97?7 107.3 5 ?I.o000 I, qo8 7 i _ .. _ ..,_.r I -.p .! q9o/.0.2 7 1.0/ 7/;C42.c,3 7 M 1, 0 .Yl-_ 7 -._' -- --.-.q 7/. 0/o 3'1.0 70l/II0,/. O44)O 1.02j/.0 7o 1.o3 7 1.o 70 q?7 I993 fI .--.-, r .o ..0.0/D 1. o 7 I. 043 1 4./[023.9q7o 990~/. rt2.' 3 I 3 I. aOfi 1. 0213 o/.757 73 27 7 1007 iO/7j13 7i 1. D/o 3 5jI 773 I / -lLog Zen PI5l//>Di.a 97t/ § /IJ/0G/6t !6l*f_ ..................- ---....--- ... .n....w.....T.....a.. -. .t...Z.. C_ ....................... _ .. 9. _tt 1.l2 7 -3L°° ,2 i l.O io7 710231 0 jl 103l 73lB /5 7c-IA 7..-..1D .l 8S PI3o63 i4to q/I-7o 1 2 t I.° 70 97 7 10-37 NO 0767(l c .2.W0D /.00 j 2 tZ /K 00 0§ 1 797l 31/°gl^f1l tt -Gol-(~ q 4i3 03 3, -7 7 17 b I -.B__ 1 1 __ls q 9II ., 1410i j$37 [9 ' 12 .j 7 / j11070 11 07 _P 4 7 j .933-O~//A v e /, I ; , ¢ ?7 [ r__/ o -.t t i ' .i7 : ' ;;.2 3 9 7 7 7 -7 1 0 1 7 f33 Il t.I ,I'__s_, .!I J 1)ri l o. I2 -.1 ' '-T ----- I- -. 'V 1 C-' IW~r~$ ____-SPUqD STRING NU MBE RS AND .LOCATIONS -177 FA CORE----I_ST wRN, 3.Q;_--J t .Ave.123 Al q.-= q Z..-qz2_qo7 [Mo° b0 i 7-- ------- -., -*F*q33 1.020 R 73 31 Ro 7 30.997/.020 I_1P3 7 Az/Ao-Jri'-I -- 4 -.------32 A?'3 I. -04xJ/1 °0 2 l i 1.o30/.o 43/. C90--~ -------I L~. ..w- V/I I'SI I f LI I _ .__ .! _. i _ " I____ __ 7 __ _0 1. 1023 7 -&7 -.0 7 I6 {,/t 7 11-7 q-., N _ __ __ _,___.q47 f.;I It3 SI e/.o6 7 1.0/ 7.7 P/o 7 t, 043-o s ---4 .a .I -C I- 1l i-a~ r,;-'. I 1.1 1. 0c0l/.G?7 1-o67 6 q,-ji3 4.q 73.I67 1.02.0 I. C'.o-.1 -j. -~ -I. _ _ ..l ,. , , ._, ...7 f , r ' , _ * .* ._ _ ,_. _,_, 7 .93PJ(.o6 7 .3 j. o 7 ({./.3 l i.7 1 3 ,oq'I 7l. 7 0 7 f i- 3 U , .l 7 7 91 3 j" l * l7 .t. .._ _ ....I 3iJ 1 d____ _ _ _ i7 ~ v /2.I I 1 0 2 1 9 ol3 I6l ,R23`i 1¢ 06731 °i 71 { 75 1.1c03 -p 713 t 1-0 57 I0 .i^7&3l/§5 °l..-~fle r. *- n .-Vt i --5 i -I 11 6 iq 7 / RA123 tCV~ 7 '"/ 7./ 0 ; (. 0 7 3 1 1 c ? 0 J f vl .D 7 l. 9 ~ S l'6 3 f .c 9 o Jl, I l ~ j 9 a a5 -i n3 .S -o £ 43 It _3. .I i.I 3 71 _ C.9?3 .' l , g3 J 2 f o 3 /. 3 } > qi0l .qy .S-.?z o. .% .._ .1 2 3 S 6 7 S 7 /01/ 12 13 t, i/ --- --7 -3..... .....-.. ...!I ..*V --?R'$ ___--..r7 3 SPN D STr R I NG NU M BE RS AN D'.LOCATIONS 177. FA CORE I l .s. j ° 5.- I / o__1 ST1RING N .1--i v*1 I I ir -t, .I _, Ii. Aw = Xf A _'.______________________ --.-.At I.1.0~3'1 (./4t 30 I. 01.,930./60_6/-a ------5 .' -.t r Th*fl-Y),023.1,c03 1.06o l (1.02oI PkC II. C'0 1. C41.3.927 5;~--..-._. -, l .tC I o13~ f1o9] X57 D J u 3 c1.o.3 1 1.6ci /;oC Ioio I .Coc0.9~,3 (.077 !*°i7 4/c-.-. _ ---0 -.-- -V I.olo/I D2-7 1.133 1.' 7.[ .70 I.o6c0 j. ()4o Zc,/.o3_3.9 33.96 o--I Pc i j BS I0 G33 1. OP 7-0 G'I 6f1'4 1/23 l -q-3 I9 9q97 w ;1 C 0, j, j-zzj i-°ooo1of .G 3 0 l,.C 1.000 lo -iP/.o;3 .3j3 S-,7 l '3 7j/.o73 q1Gl/3 l/_ _ _ _ __71. t _~ 2._7 q N71°7 l 993 11°07 I/.G/ 7 lI2ey06?'0l85T /- -Dl£; I-,-~, I -II 7I3 _ G 7 _ _ /.7 0ll5-__ _ _ [ 3 _&.46..1 1 > 'i11 7 a 7. v- .,7 J[7 /c 7 a)3 lA -1?c.92 Av °__ 7 7 101 /1 J7 /37/13 2 I ... .....

* ..C w t----------------

SP14D STRING NUMBERS AND ' STRNCG Nr LOCATI-E S 177 FA CORE' I U A* I.=I.- I w l',*~ olo49 %\ 3.ql3 W1°/° l/,020l S-.% 1 1 1 31i 30 -iA..4L77 f.o 0 2 3 >964/.D20 l .997(?3 V/ 7o7 ko? _ /I. o /t .0-0 5- /-D/ 1. -V, /. 02 G / 1. -Y *1. 1030 / -/0-(.033 /,,60 j.023 l/ 7G o G 777 1. 27 f .07o /1.7 I.o? 1.oo3 / 03 S7 /co 02 7 1.0/c 1. 63 1 fc~o j.f00 j(097 (0&qoj(.o63 J 063 1.030j /23 /7 /7 I6o°t47 0to7 /31 I. 090 1/.t2 0 //qw7 00/.0 6 0 .t_3 0.7-3 I. C&7l 1.0g3 1 .o,1~- ({.OEg / a f/ -.- *WS, S 41 ta A .t.,1n j-fr -*; 7 1 VZOf°l 1-077 l C13jI ,.°Soj / .0S3 1.077 1h037 /1.00o /,04o 3 / 53 /.0/7 /7 I o* (0/0(o2(l7 j9 27 070 (.03- fgc3 jO-Lf 7 A/37 ' 3I 33I 1Io 7* I-- --I .! li___ o____&O 3lc3/o~i/o /o7 /.o Blo7lc.003 I ( 13 1173 o7 I I, -I I T 1 I 1/o6O o / 'Oz 7EO r, p' 7 3 7 I 013-P03 ¢0?7 / C§JI?/.12 03 1.0YPI .O3 9,ib, , 7fi 15J 9,k-iOI -_ _7 7 M.770 7 l I-7 6 7l 7V 7 .7 ...7 1 3 9M 5-6 7 7 10/ / -2 /3 Hz.___._ __. _ .P T I ...*t~tIgt ' l to *,9_ _ _ _ n .1 ._.- ._.. -._ .,_ __*~ r- .-t\;<5* F 5 X .: -.- ------Br~76~oF(cr I417'~o~J WISTPA~ ,.-, ,_*~6m = f\ -1.'-i iI:% ., 3 9?3" I t 3q63.0 1*1!p 4 i 71 42/ 2 /.02 Z].I% ..1.t. .__ '.. S 2 S.. ..I ..7. ..I.!f*,.,..I/ .I.t I.II ": 1.38a:. ./,G6 I 1/.00 31 , W1 30, 1.020 1 b/.o/7./.00/.CzZ.-l. ..1. 041 2.- .-=, __ __ _ .1.0 I)SI 2 kozo/.0ss I/. c 4 1.06;/ L/.°043S.'.-.--. -* -- I

ctrr. rvs, n,.-."/3/,0 61 1.?bI/.070 1.0 73 1.0/8/-/ 717/ I q;-2 ,_..--._ _AVe :1C-I.oaq 11 OF i/. 67r t.o63/o 79/. 045 ZG/. IIA 1 0o3 2.-.4 ..4 ---. 1 -- --- !- --r----I '-.1/l .04[.1;7 1t3 0 1 (.0(141 11.08c5 10o3 1/07q 1I2 A I/. 6;5--.-IXo_ I S I -I1. o 1 -.o 1Z I
__-,`_ -l 5 7 0 10aZ7 f.o6l foi/3jA 1?67 [. £77 j q10" 7.a r r 3.3 .-..- .. 1 s _..*+ / i/ 10 ,os7 li .l L.70 1'.o33 1/3, i.~ofiil./o'o l/-o37 J t0.D j O 1/.c7 IsS-7 7 -, 0- 7 7l4 -e.- - -77 o -I /.c---* o03/.o13.517 o67/.o67 1.037/. O4!3 I. 04/3.IG 1.013 1.og.` 77 1. -/3'-9t? 7 J -l .,i15 i -s-v D3 .C- / 33 93!,87 93O iAtlAt 1 .7 .96 ,9~ _ ; _7.27£1?I.770 , R 73 WI'-7-7.7/7.7& 7._ _... .._._/hi A VC = .7 7--I , Al ~.770.8' 7"cil-~ Eil£§ S 7 7q7l7 live = .3cS^U .I;.I1 I 7, I I 8 7 .v/ 0 l I /1 3 1'I ._ _ .-1 i .

-)E suRv., & GE WINLCOX COUP, ..-ie. ).WER GENIERA-51011GP.OUP J. D. CARLTON.From -~M~~L I.P. S. M\RTELLS 3. R. BURRIS 3. 11. KNOLL P. A. TREVE:;TI J. R. GLOUDEMAS C. E. PARKS£ A DqwTha D Pi LELE t cDS eW Cust 14-2 *.. sFile No.TMI-2 or Ref...b .Date.. CORE BYPASS FLOW FOR CORE BLOCKED AT TOP ONLY APRIL 8, 1979 alfbh lobt Ito coer one gugvea-r glnd *s subje 1 ct onJ 7.*

REFERENCE:

1EO.£0 R. M. GRIBBLE TO J. D. CARLTON, -"CORE BYPASS FLOW THROUGH CORE BASKET (UNIFO&IE BLOCKAGE) 4/8/79.THE REFERENCED MEMO REPORTS CORE BYPASS FLOW FOR A UtIIFORMLY BLOCKED* CORE. F.OR THIS CONDITION, 22% OF THE VESSEL FLOW BYPASSES THE CORE.THROUGH THE CORE BASKET. ANOTHER CASE HAS BEEN CONSIDERED AND IS* REPORTED HEREIN.)*CORE BLOCKAGE AT ONLY THE TOP OF THE CORE HkAS BEEN ANALYZED TO DEIERMINE* CORE BYPASS FLOW FOR-TIIS 'MORE LIMITING SITUATION. RESULTSOF THIS.-ANALYSIS INDICATED THAT CORE BYPASS FLOW FOR THE CORE BLOCKED ABOVE.THE UPPERMOST INTERMEDIATE SPACER GRID WILL BE APPROXMLATELY 27% LEAVING 73% AVAILABLE FOR CORE HEAT MEIOVAL.,.I , ., .* '. .' .,. ., '. ' ...MAJOR ANALYSIS ASSUMPTIONS FOLLOW: ....1. MAXIMUM RESISTANCE OF THE CORE AND CORE BASKET-83 X 10 8 PSI!(LB/SEC)Y, 17.7 PSI AT 4600 LB/SEC '.2. NOMINAL CORE BASKET RESISTANCES

LOCA HOLES (CROSSFLOW)

R 7.64 X 10 PSI/(LB/SEC) 10-6/E)*UPPER BASKET R 5.43 X 10 PSI/(LB/SEC) 2 3. CORE CEOKETRY IS NOMINAL BELOW BLOCKAGE I -J.1D. CARLTON I , .--2~-APRIL 8, 4. CORE BASKET GEOMETRY IS UNDISTORTED.

5. RELATIVE RESISTANCES OF THE CORE AND CORE BASKET Ra4AINl UNCHANGED DURING NATURAL CIRCULATION CONDITIONS COMPARED TO THEIR VALUES DURING 1/O.PUNP OPERATION..

a ..: IA.W '64 .4 .'..4r!; ; -.,<A) ,. !.,-..I-e.,* .. .- * .7 iA..* --

1!LE 'BABCOCK & WILCOX COMPAIJY)OWIE GE1'ERAT ION GROUP'o .G. A. MEYER, MANAGER, T-11 ENGINEERING UNIT!rom rlvA k i .THE-79-194 ZDS 663.5._ ...j,7 --._ , -J. A. WEIMER, T-H ENGINEERING UNIT, EXT. 3236"u 5t. M- FulIe 11o..TI -2 or .Ref.:ubj. .Date INCORE THERMOCOUPLE ERROR EVALUATION APRIL 10, 1979 T hi Ialr ea o wOar on* .cviomr ond *n. subjet only, AN ANALYSIS WAS DONE TO DETERIMINE THE MAGNITUDE OF INCORE THEDIOCOUPLE ERRORS FOR TMI-2 PRIOR TO MARCH 28, 1979. THIS ANALYSIS WAS BASED ON A TEIPERATURE AND POIYER DISTRIBUTION AT 98%AND. 15% FULL POWER. THIS WORK ASSUMES THAT THE INLET AND OUTLET RTD (RESISTANCE TEMPEPATURE DETECTOR) TEMPERATURES AND POlYER DISTRIBUTIONS WERE CORRECT, AND INADDITION, ASSUMED A CONSERVATIVE + 3°F DIFFERENCE BETWEEN THE CORE OUTLET AND VESSEL OUTLET TEMPEPURTURE A 98% POWER. THIS RESULTS IN A 0.5 0 F DIFFERENCE AT 16%o POWER. MORE REALISTIC TEMPERATURE DIFFERENCES (IE. 22 0 F AT 98% FP AND .2 0 F AT 16% FP) WOULD INCREASE Tl-IE PREDICTED T-C ERRORS SLIGHTLY.THE METHOD USED FOR THIS ANALYSIS WAS BASED ON A KNOWN BUNDLE.DELTA ENTHALPY, AND FLOW RATES (FROM ONLINE COMPJTER (OLC)) FO'R.AN AVERAGE POWER BUNDLE (RELATIVE POWER -1.0). THE EQUATION USED FOR THIS ANALYSIS IS: Q2 l x (H -HIN) + 1 O 1 UT 2 HUI IIIN1 a s.n G. A. MEYER; APRIL 10, 1979 PAGE 2* WHERE RELATIVE POWER OF BUNDLE FOR EACH CALCULATION (FROM OLC)Q, = RELATIVE POWER OF BUNDLE FOR AN RPD OF 1.0 Q1=1.0*.W1 BUNDLE FLOW FOR AN RPD OF 1.0 (FROM OLC)W= BUNDLE FLOW OF BUNDLE FOR EACH CALCULATION (FROM OLC)HOUT H DELTA ENTHALPY FOR AN RPD OF 1.0 1U IN 1 ET RPD -RELATIVE POWER DIFFERENCE (NOR14ALIZED TO AVERAGE ASSEMBLY P-OWER)1 H UT CALCULATED BUNDLE OUTLET. ENTHALPY FOR EACH BUNDLE.I .I ..C ..* HOUT IS THEN CONVERTED TO T AND COMPARED TO THE MEASURED TOUT 2.(T-C READING). THIS ANALYSIS (AT 98% AND 16% FP) WAS EXTRAPOLATED TO 1%FP. * -ANY INHERENT ERRORS ON THE OLC FLOW AND RPD CALCULATIONS ARE* ELIMINATED BY THIS RATIOING METHOD. THEREFORE, THE ONLY REAL -* UNCERTAINTY IS IN .THE HOUT AND HIN MEASUREMENTS. THESE WERE* ASSUMED CORRECT FOR THIS ANALYSIS.THE RESULTS OF THIS ANALYSIS INDICATE AN AVERAGE + 7.94 0 F ERROR AT 98%, AND A 4 s.59 0 F ERROR AT 16% POWER. THIS EXTRAPOLATES TO A + 5.16 0 F ERROR AT 1% POWER. 4.G. A. MEYER APRIL 10, 1979 PAGE 3 i ., ASSUMING NO DAMAGE OCCURRED TO THE T-C'S DURING THE TRANSIENT OF MARCH 28, 1979 AT TMI-2 THESE RESULTS WOULD APPLY TO THE PRESENT T-C READINGS, THUS IT IS POSSIBLE THAT TIlE INCORE ThIERIOCOUIPLE READINGS PRESENTLY BEING OBTAINED ARE HIGH BY AN AVERAGE OF 5 0 F.FINALLY, THE AVERAGE T-C ERRORS WERE. CALCULATED.AS A FUNCTION OF DIFFERENT POSITIONS IN THE CORE. THE RESULTS SHIOW1 NO INHERENT CORE REGION DEPENDENCY. i JAW! SGH*.CC: FE UNIT MANAGERS.J. S. TULENKO CORE--HOT SPOT TASK FORCE QA: THE METIHOD AND CALCULATIONS

WERE REVIEWED AND FOUND TO BE CORRECT AND CONSISTEN'T WITH THE STATED ASSUMPTIONS.

o rDATE -°#I. PCT d. ../ _ /el.el. ~6 wIwiJ"3%77 I _---- .i0 Al 1'4s A AT /iVQ. a, 6 C -*'t7...L7-/1- --_-~t2442pCJLuk.

.

.10.. .. ..# .. ji -'- w A t_ 'J, , ,2>,,..S, .<) 0.. ._. .......... .................. ...__..........._.....__ ......... .................. ... ... ....... .......Vr.u, .V-L.,v -'A'.__._ _ _______ _ ___ ______ _, __.___.___ _ ___ --n At-e- ----Z-~~___,,. .__ ...... -.ll ---k' I--..-zl T3l---cv.. -- cllv-fj -._ _ ). u I. ___________________________________________ T/c TrJ-y11.9 d31 Tl- -F~~&i~iJ~i~\4Lv .--.----.-- PI, LT* A ,-...6;Z10 'j >. Q 70.1V.lC I '.., I 4 v ..aI 0 U.ow., THE BABCOCK & WILCOX COMPANY .CWEiR GENERATION GROUP 'K F *, lq 'AfA -G.A. MEYER -MANAGER, THERMAL HYDRAULIC ENGINEERING From R.M. HIATT -THERMAL HYDRAULIC ENGINEERING THIE-79-191 $DS 663M Cust. File No.or Ref.TMI-2.Subj. Date LYNX1 MODEL FOR Tia-2 BLOCKAGE STUDY I APRIL 10, 1979 lbi lett er to co-r e a- , o ar d one subject only.THE OBJECTIVE OF THIS WORK WAS TO DEVELOP A METHOD FOR ANALYZING LOCAL COOL-ANT CONDITIONS FOR TkII-2 DURING SELECTED TIMES OF THE RECENT ACCIDENT. AN EQUALLY IMPORTANT CONSIDERATION WAS THE DETEPRMNATION OF LOCAL COOLING CAPABILITY FOR THE BLOCKED CORE UNDER NATURAL CIRCULATION. FIGURE 1 ILLUSTRATES THE NODING SCHDIE FOR THE SIMPLIFIED CORE MODEL.THE TWELVE CHANNELS WERE DESIGNED TO SEGIENT THE CORE IliTO AREAS THAT IT: .WAS BELIEVED EXPERIENCED IAJOR DAMAGE, SOME DAMAGE AND POSSIBLY NO DAMAGE.* LYNXI PERFORMED WELL WITH THE MODEL FOR A CLEAN COiPX. HOWEVER, WE WERE NOT ABLE TO ACHIEVE THE HIGH PRESSURE DROP (18 PSI) ACROSS T1UE CORE WITH TRE FLOW RATE FOR ONE PUMP OPERATION ANb A BLOCKED CORE CONDITION* PREDICTED BY CONTROL ANALYSIS. THEY PREDICTED A.FLOW? OF 4500 LB/SEC FOR CORE PLUS CORE BARREL-CORE.BAFFLE hNNULUS FOR A BLOCKED CORE CONFIGURATION1. SINCE LYNX1 STRUGGLED FOR CONVERGENCE DUE TO A SENSITIVITY OF THE CODE TO THE BLOCKAGE MODEL, A NUMBER OF MODELING SCHEMIES WERE TRIED WITH VARYING DEGREES OF SUCCESS. SEVERE BLOCKAGES AT EACH SPACER GRID (K=30-35), ANUl INCREASE IN WETTED PERIMETER FOR ALL C1LNELS, AND SLIGHT VARIATION IN* RESISTANCE.FROM CHANNEL TO CHANNEL WERE ITERATING PARAMETERS. THE BEST ESTIMATE OF LOCAL FLOW1 BEIIAVIONs OUTAINE)D TO DATE ACHIEVED X1 UNRECOERA!3LE -1 ... .I PRESSURE DROP OF ABOUT 7 PSI FOR 4722 -LB/SEC CORE FLOW FROM CONTROL ANALYSIS AT-PRESENT IS 18 PSI FOR 4500 LB, CORE PLUS CORE BARREL-CORE BAFFLE ANNULUS.ITHE BEST ESTIMATE'SEC FLOW THROUGHTHE A CHATA CASE WAS MODELED TO GIVE THE FLOW SPLIT BETWEEN THE CORE AND THE CORE BARREL-CORE BAFFLE ANNULUS. IT WAS ESTIMATED .THAT 78% OF THE 4500 LB/SEC PREDICTED BY CONTROL ANALYSIS FLOWS THROUGH THE CORE WHICH AGREES WELL WITH CONTROL ANALYSIS ESTIMATES. THUS, BEST ESTIMATES TO DATE SHOW THAT 3510 LB/SEC IS FLOW4ING THROUGH THE CORE WITH 1 PlrIP OPERATION. ONa RIPORTANT FACT THAT WAS EVIDENT FROM AN ENERGY BALANCE ON THIS FLOW RATE IS THAT THE INDICATED THERMOCOUPLE AT IS NOT POSSIBLE CONSIDERING A 4-5 IF.4 DEC.T HEAT RATE UNLESS THE THERMOCOUPLES ARE MEASURING LOCAL EFFECTS, SUCH AS AGGLOMERATIONS OF PELLETS NEAR THE THERMOCOUPLES. THIS APPEARS TO DISCREDIT TUE THERMOCOUPLES, THEREFORE, SOME DISCRETION IS NECESSARY IN THE INTERPRE-TATION OF THIS DATA.E AlY IN CONCLUSION, ALTHOUGH LYNX1 MODELING HAS NOT BEEN SUCCESSFUL IN MATCHING FLOW AND EXIT PRESSURE AT THIS TIME FOR A BLOCKED CORE WITH ONE PrMP OPERATION; IT IS BELIEVED.THAT AN ACCEPTABLE MODE CAN BE DEVELOPED. THE ADVISABILITY OF ADDITIONAL WORK IN THIS AREA DEPENDS ON THE WORlM SCOPE OF-FUTURE WORK ON THE TMa-2 ACCIDENT. FROM PAST EXPERIEACE, THE MODEL* DEVELOPIIENT WILL NOT BE QUICK BUT COULD REQUIRE A MONTH'S EFFORT.R 11H/FFA CC: F.E. UNIT MGRS..X'J.S. TULENKO CORE HOT SPOT TASK FORCE ft .I ..7.. Ai1 0.)10..I I B = t F2- "fl 1.2-gld'v l I / , J -IUo (. ....O'Kill)) I ---/, i -, THE BABCOCK & WILCOX COWPANY POWER GE'IERAT ION GROUP T1%0 TIJE-79-190 To I G. A. MEYER, MANAGER, T-11 ENGINEERING UNIT P. J. HENNINGSON.. T-H ENGINEERING UNIT, EXT. 3515 12E12 .13* EDS 663.5 Cust. -.File Ho. 86-1100502-00 or Ref.TMI-2 .660-021A Subj. Date DAMAGE MODEL -FLUIDIZED BED APRIL 10, 3979 I i. *.-..*Cd ** uic n I. .. .-..------ ---.- ...- ---- ---* I

REFERENCES:

1) MEMO, CORE CONDITION TASK FORCE TO J. S. TULENKO,"CURRENT ASSESSMENT OF CORE CONDITION, APRIL 7, 1979 (1800)," APRIL 7, 1979 (7:48 PM)-2) PERRY'S CHEMICAL ENGINEERS HANDBOOK, FOURTH EDISON,.PP. 549 -551.3) MIEMO, P. J. HENNINGSON TO G. A. MEYER, "POSSIBLE MODE OF INCREASED T.C. READINGS," APRIL 7, 1979..THE DAMIAGED TMII-2 CORE WAS HYDRAULICALLY .MODELED AS A PACKED BED.THE MECHANIS'M OF FUEL FAILURE WOULD RESULT IN.APPROXIMATELY THIS GEOMETRY AND BE LOCATED IN THE UPPER REGION OF THE CORE.BRIEFLY THE CORE WOULD BE CONFIGURED AS UNDAMAGED FUEL UP TO A HEIGHT WITH DAMIAGED FUEL (FUEL PARTICULATES AND CLADDING)

ABOVE THIS RESEIBLING A POROUS MASS.*THE BASIC CONFIGURATION OF THE CORE WAS OBTAINED.FROM REFERENCE 1..THE CORE WAS ASSUMED UNDMIAGED AT THE PERIPHERY WITH INCREASING FAILURE TOWARDS THE CENTER. PARTICLES OF FAILED FUEL WHICH COMPRISED THlE FLUIDIZED BED WERE ASSUMED TO EVOLVE FROM THE FOLLOWING FAILURE MECHIAN ISM: I I C. A. MEYER APRIL 10, 1979--- .PAGE 2 THE FUEL CRACKED ALONG TWO PERPENDICULAR AXES LENGTHWISE AND ALONG THREE AZ$ PERPENDICULAR TO ITS AXIS.-THE MASS OF FUEL WOULD THEN CONSTITUTE THE MAJORITY OF THE CONGLOMERATE WITH CLADDING FRAGMENTS ASSUMED TO HAVE A SIMILAR.GEOMETRY. A .SUITABLE CORRELATION FOR PRESSURE DROP THROUGH A PACKED BED WAS OBTAINED FROIM REFERENCE

2. THIS CORRELATION (ATTRIBUTED TO LEVA)IVAS APPLICABLE IN THE HIGH REYNOLDS NUMBER RANGE EXISTING IN THE DAMAGED CORE (RE X 10,000 ). IT IS IMPORTANT THEAT THE RANGE OF REYNOLD'S NUMBER APPLICABILITY BE ASCERTAINED FOR A GIVEN CORRELATION.

To THE SENSITIVITY OF THE.FRICTION FACTOR ag CHANGES IN FLOW FROM VISCOUS TO TURBULENT IN THE PACKED BED CANNOT BE NEGLECTED. AN ATTEMPT TO MODEL THE CORE AS DEFINED IN REFERENCE 1 WAS MADE.THE LOW RESISTANCE IN THE PERIPHERAL BUNDLE CAUSED THIS METHOD TO FAIL. IT WAS THEN ASSUMED THAT FAILED FUEL (OR A CONGLOMERATE OF PARTICLES) EXISTED AT THE PERIPHERY; THE CORE TOOK ON THE FOLLO'1.'ING SHAPE: CENTRAL BUNDLES (116) 4 FEET OF FAILED FUEL BELOW THE UPPER END FITTING (fAcKEDb BEZ REMAINING BUNDLES 2 FEET OF FAILED FUEL BELOW TIlE UPPER END FITTING

*-G. A. MEYER APRIL 10, 1979 PAGE 3 THE GENERAL SHAPE' AND RECOMMENDATION.OF A FOUR FOOT HEIGHT WAS OBTAINED FROM REFERENCE 1.A TRIAL AND ERROR, APPROACH WAS USED. THE VOID FRACTION OF THE PACKED* BED WAS VARIED AND A CORE AP CALCULATED.

THE FINAL RESULT WAS THAT FOR THE ABOVE CONFIGURATION A CORE AP = 14 PSI WAS OBTAINED FOR A CORE FLOW OF 13.1 x.10 6 LBM/HR. THE FLOW IN THE CENTRAL BUNDLES (61) WAS .058 x 1o6 LBM/HR AND IN THE OUTER BUNDLES.082.x 10 6 LBMI/HR. THIS WAS FOR A PACKED BED HEIGHT OF FOUR FEET* AT THE CENTER 61 BUNDLES AND TWYO .FEET ON THE REMAINDER OF THE CORE.A VOID OF 60% WAS USED WHICH COMPARED WELL WITH THE 50% RECOMMiENDED IN REFERENCE 1.NO FURTHER ATTEMPTS WERE MADE. TO MATCH PRESENTLY ASSUMED CORE CONDITIONS -AP ',16.0 PSI, CORE FLOW "l 14.106 LBMI/HR. VARIOUS ASSUMPTIONS CAN* BE MADE CONCERNING THE GEOMETRY AND MAKEUP OF THE FAILED FUEL WHICH IS ASSUMEI) TO RESEMBLE A PACKED BED. WHAT IS IMPORTANT I-S* THAT:* 1) CORE CONDITIONS COULD BE APPROXIMATED WITH THE PACKED BED. ASSUMPTION, 2) FAILED FUEL (OR A HIGH RESISTANCE EXISTS ACROSS THE CORE).THE FUEL AT THE PERIPHERY COULD BE UNDAMAGED WITH A LAYER OF PARTICULATES BENEATH THE CORE SUPPORT PLATE ALTIIOUGll IT SEEMS UNLIKELY THAT THE MATERIAL WOULD BE THAT NON-' HOMOGENEOUS. . G. A. MYER APRIL 10, 1979: PAGE 4 FURTHERMIORE, IF THE FAILURE MODE OF THE CORE DESCRIBED IN REFERENCE 1 IS ASSUMED, THEN IT APPEARS THAT THE THERMOCOUPLES COULD BE SURROUNDED BY UO 2.THIS WOULD EXPLAIN THEIR HIGH READINGS. THE EFFECT OF UO 2 SURROUNDING THE THERMOCOUPLE WELL .AAS DESCRIBED IN REFERENCE 3.*PJH/sgh CC-: i; H. JONES-'A. B. JACKSON J. C. MOXLEY* D. V. DEMARS B. J. BUESCHER* .If. WILSON R. A.. KING D. C. SCHLUDERBERG-G. S. CLEVINGER QA: THE METHODS PRESENTED HAVE BEEN REVIEWED FOR APPLICABILITY AND THE CALCULATIONS SPOT-C.HECKED FOR ACCURACY AND CONSISTENCY. THE METHOD IS DEEMED APPROPRIATE FOR THIS-PARTICULAR APPLICATION.

Jc/4e< Iw. 2'7IkPPMCTE___{__f TI-E dABCOCK & WILCOX COMPANY.6~rl ` i -n vr n A -r ir j -n ir i P rvricli i..liCr[mI Ia WI's .IEWJUul To 1 -,Z J; I .ANDREW'S ...../.-'IF.From.COR. COMNITION TASK FORCE Ont &A91 t Cust. File No.or Ref.subj .Date~ESTIMATE OF LOOSE CORE DEBRIS VOLUM.E (4/9/79 -2000) TIME._ ._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ A P R I L 9 , 1 9 7 9 i ,is lot. so covr .e. cuslemee and one subject otfly.I ATTAQ 1ED IS AN ESTIMATE OF THE AMOUNT OF DEBRIS AVAILABLE FOR CORE BLOCKAGE AND ITS POTENTIAL DISTRIBUTION.

MIJS IS TO AID IN THE OVERALL ASSESSMENT OF CORE BLOCKAGE.W,)bST OF o iE I-NTOrtTION IN FORMING -MIS ASSESSME3NT' IS STILL PRELIMIMMAY

AJND IS BASED ON OUR BEST ESTDIATE OF CORE nAMl\GE.CC: P. HEMNINGSON CORE CONDITION TASK FORCE .. -J. S. IJLENTKO lATTACHMENT CORE BIPCKAGE ESTIMATE BELOa IS AN ESTIM'.ATE OF THE LOOSE MATERIAL AVAILABLE

, BLOCKAGE.O CONTRIBUTE TO CORE 1.ASSLME 30% OF TOTAL Zr CORE INVENTORY IS OXIDIZED PRODUCING ZrO 2 45,000 LB TOTAL FUEL CLAD INVENTORY 7,900 LB OTHER Zr IINVE-ORY 52,900 LB TOTAL O 15870 LB OF Zr IS OXIDIZED WITH A 1.6 BULK VOLUME INCREASE, FORMING 63 FT 3 OF Zr.THE TOP 30% (46 IN.) OF THE CLAD OXIDIZES EXPOSING 41" OF U0 2 TO T1lE COOLANT.a TOTAL U0 2 EXPOSED IS 57,400 LBS OR 94 FT .O ASSUME AN ADDITIONAL 10 Fr 3 OF MALTERIAL IS EXPOSED FROM FUEL ROD (SPRINGS, ENTD PLUGS, ETC.).2.3. BASED ON 'IHE ABOVE THE EXPOSED Zr°2* U°2 OTIER*4. OF THIS, SOME IS CAPABLE OF BE]VOLUME I AVAILABLE ZrO 2 63 U0 2...94* 167 10* 167 MATERIAL AVAILABLE FOR CORE BLOCKAGE IS: 63 FT3 _94 FT 3 3 167 FT ING MOVED BY FLOW T3 MOBILE 45 (FLAKES 14 (<1/16" 59 -& DUST)SIZE)INJOBILE 18 LARGER FL'\KES Cl.ON RODS 80 (>1/16")10 108 THE SMtLLER PARTICLES IMY EXIT ND AMVE WVIlH THE FLOWMIND MAY RE-DEPOSIT IN THE CORE OR SETTLE OUT ELSEW1HERE IN THE SYSTE4I.EQUIVA1NT-FLUY BLOCKAGE o ASSUME THE EQUIVALENT CORE FLOW AREA (10.6 FT DIA), IS 88 Fr 2_ ..1 o TOTAL EQUIVALEr DEPTH OF BLOCKAGE IS 167 = 1.90 FT ASSUMING SOLID MATERIAL 'o ASSME 1.5 VOLUME INCREASE FOR PACKINGTHEN THE EQUIVALEa-r TOTAL DEPTH IS 1.90c 1.5 2.85 FT o IT IS EXPECTED THAT THE AM-IAGE- WILL BE GREATER AT THE CEWTER TH{M AT THE CORE PERIPHERY (SEE N\ECT SECTIO\N). CORE DAWAGE DISTRIBUWION 1lE CORE DANAGE WILL BE ORE SEVERE IN TIE CENTER OF ITE COPE TILkN ON THE PERIPIHY. TIS RESULTS FROMl THE CORE DECAY HEAT POIWER DISTRIBUTION 1W-IIfIG CLOSELY FOLLOl'WS THE CORE POWER DISTRIBUrION PRIOR TO SHTDOIWN (SEE .G. 1).IHIS WILL RESULT IN CORE. DANGE DISTRIBUTION AS SIHOW1N IN FIGURE 2. THE FUEL RODS IN'PERIPHERU ASSDSBLIES 1LkY BE RELATIVELY INTAC UHILE THE CEIWER ASSEMBLY IS PROBABLY SEVERELY R1kNMGED, POSSIBLY TO THE CENTER OF 7TE CORE. THE CETER ASSEMBLIES MAY HAVE VIRTULLY NO RECOGNIZABLE ARRAY IN THE UPPERMOST GIID SPANS.I

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I __________________________ L4 :1* N7f___________________________________ 77] __--it-;:;-- 71 HE-BA-BCOCK&-WILCOX COMPANY POIQMER GEI-ERATION GRO'UP'1 L l N To -DISSTRIlUTION -- -- -- -From P. S. BARTELLS, TECHNICAL .STAFF.tDS 6.'* Cust.': Subj.TMI -2 2 ii 4J I File No.or Ref.CORE PRESSURE DROP FOR NATURAL CIRCULATION CA.rCTTI.ATTON\ Date APRIL 9, 1979 this , flor to eto-r an* tvslomer ond 9Ao subjeit only.DISTRIBUTION B.A.C.E.R.B.B. M..J..s.M.R.B.A.Womack Parks Davis Dunn Tulenko Gudorf'arrasch C,D.B.E.G.F.J.J.R.H.J.D, J.A.Morgan Bingham Malan*Cudlin Stoudt Carlton Castanes-' ..H D .H. Roy E.F. Dowling.*..RHM -Hiatt J.M1. Knoll P.A. Trevent.:M,. AtRIL Knoll* 4 .. ....

REFERENCE:

MtEMO, SAM~E SUBJECT,. APRIL 6, 1979.i ADDITIONAL. INFORMATION HAS BEEN OBTAINED FROM TMI W',rHICH FURTHERt SUBSTANTIATES THE CONCLUSIONS PRESENTED IN THE REFERENCED MEVO.BASED ON-THIS INFORNLATION, AND THE FLOW SPLITS PREDICTEDEY THE-PUmP. COnE; AN 18 PSIA DROP ACROSS VESSEL DOWNCOMER AND CORE (AT PRESENT CONDITIONS AND CORE FLOW OF 4500 lbm/sec) IS IN'DICATED. &ca 4?7' C, LV'.A ATTACHMENT 1 IS THE TABULATION OF RC LOOP FLOW TRANSSMITTER DIFFERE.NTI.'. PRESSURE SIGNALS OBTAINED AT 4:00 a.m., APRIL 9. ATTACHMENT 2 IS THE CONVERSION OF THE TRANSMITTER VOLTAGE MEASUREMENTS TO LOOP FLO;S.AS A FURTHER CHECK ON THE RESULTS, I HAVE ASKED MIKE KNOLL OP CONTROL ANALYSIS TO ANALYZE THE SAMNE CASE USING THE SPLIT CODE.IT IS HOPED THIS INFORMIATION WILL BE AVAILABLE BY LATER TOMOR-rIOW AFTERNOON. AT THIS POINT -I WOULD.LIKE TO ACKNOWLEDGE THE EFFORTS OF MITKE KNOLL AND PHIL TREVENTI IN PERFORIMING THiIIS ANALYSIS (WITHOUT WI11CH I W%.OULD STILL BE SETTING UP THIE INPUT).IF ANY ADDITIONAL INFORMALTION ON PLANT STATUS (WHICH COULD PROVIDE A FURTHER CHECK ON THESE PREDICTIONS) IS KNOWN TO BE AVAILABi.E, PLEASE CONTACT \1E1 IMMlEDIATELY. PSB/DiH G3~-~~ O(S\ .2 -tjz./- 6 .ij~ .\~. w ..t-i ..A('n.j I , , , A .I clkk-6-E: Q~., v,(rV.I I .06' 1 0.-Et 0 to I9\yn. / tl,.2. 10 Q q4.%.j .a~1kJ~.El a...'.- 7. VW, .i fc;. t .\ OQ.I..-Q Ck,,1.., .. U...Q..........,, WN.-.:........ .* * ., * .. .aL _ ..cs........ ..- : ..Y _.. .. -.Ck.... ... -e ..... ..,.- w .....w ...Atv s Qa1Ag.are 3 UX'S.3 k4~Qt' .=. LO.04- --t.A- 31A,..4 -, 4 (. --V2-"-J. L Sc&A5- .- g .V A'I C.4 n- \f 'U .... .. .. .-... .. .I .. .- C.3 Z(.W. V ~(Z, oe Z ..C.C- b 3 ,- ) _.02 o Z V i A .r v\j~ 2ao X'.0 .I (4.J"L 9 y d ' % ./ 5. ... ., -....,. I,-I-I I I lIX ...I* .^ * + {BS\ .-P -.r^-,- ...j ._.1 ..-I .@.@. I.,%. I Je.- O' ).' " ft I__ _ (__-_ \ ,Z II I a I(91 e'I i t .k: C)vS&&,,.¶a i o) A.~tS .S Wo .= *-WA I .I w .d a H-~mr~/I Z- 58. o ( !'-z -3 1 13:?..At-/GZI23 2kr6A /0.f 3 4 l~-~S a u 0.1= 1 .T-I AJerPi* I I v Rcst e L sea:Ki~eA-n 4)hei&4 f q9 t- -L Ard a'I --@-,, -.I -.Ix;. 14 ^r (7-1, 0.Aa* Kb (.* ( It s )bi .17 .9 ol1S1-k..__b __,,_ _.I 1¢g.Q?,ciz.I?1 si 9\ !a '7 I I 151"S",q7 161 r;..V c LEG c- S A l:--l?II.\1 .Lp -p C', t;L9.As w ;;.c>:.A I ....., .I 9. C, I*a %*-._ , l tt~*ka I (O 0wv)9'-48 go ss8'I,-C.o19 1.4.L 0 .p is..._ 4 a)kV~c.'7- :Iquotas.JI'3 ' Igl' -I I' knJ THE StABCOCF( & WILCOX CC' 'PAIJ.POWlER GEIJERATION GROUP To .l, .o D I ST R B U T I O 1 From..A I Rtoc A4 Ps UAKILLLb, ILeHNILJAL ZI-r * .-Cust. -' ' File No.i CORE PRESSURE DROP FOR NATURAL CIRCULATION Date CALCULATIONS APRIL 6, 1979 10:10 PM Uh. falls# to £*O.G t stom*t @ofid ot sibjetf only.I D I S T R I B U T I O N EA .W10.ACK DH ROY CE PARKS BE BINGHMM1 RB DAVIS JD CARLTON* ...81M DUNs'N JS TULENKO.MR .GUDORF.BA KARRASCH JA CASTANES CD IORGAN... .GF MALANI*^!. , *- JJ CUDLIJY I ...." .... ..An analysis (using the PUMP code) was performed earlier this week to estirate core flow blockage. Vensel (i.e. core and bypass) flow resistance was yaried over a wide range and the change in loop flow rates, core flow rates and vessel delta P..were calculated. The results are tabulated.below: (* means unblocked core)Rv* 1.712** .3.5.7.0 10.0 .15.0 30.0 60.0.1.6*2.98 5.05 6.45 8.22 11.89-15.79 WV 1017 7Ok 9450* 8596 .8095 7443 6291 .5136 WHLA.12810*12690 12500 12380 12200 11870 11510 WHLB-2643**-3244-3900-4281-4752-5580-6376 Where: Ry -downcomer

core + bypass flow resistance, (psia)/(lbm/sec) 2 X 108 aPv pressure drop across core + bypass, psia WY -core + bypass flow rate, ibm/sec WHLA = hot leg flow rate, active loop, lbm/sec WILB- hot leg flow rate, idle loop, lbm/sec

*i .J. IUI.*r ._I/in be seen from the tabulated results, the active loop flow is not a* ong function of the vessel resistance.

This is due *to the high reverse/ows through the idle pumps. However, the reverse flow through the idle loop is a strong function of the Vessel resistance. Prior to' this afternoon,/ I had been under the impression that no method existed for calculation of reverse flow in the idle loop. Recent information from the I&C group shoves this not the case. As early as last weekend they'estimated the reverse flow to be---14.5% which translates to -6444 lb/sec. A further check today results in an estimate of -6797 lb/sec. Based on the tabulated data, the-vessel pressure drop is at least 16 psia.Additional evidence-to back up this is the indicated flow in the active loop Which is consistently indicating 49-50% of nominal which translates to approx.-'11,000 lbin/sec.-' 'llOO~bm/se. .... >._.Separate calculations by Jim Veenstra and Larry Losh (see attachments) on 4/4/79 place measured flow in the active, loop at 88,350 GPI (based on Gentille, delta P -173"), which is a flow rate of 11,445 lbm/sec.' The attached figures indicate that the core and bypass pressure drop is between 16.7 and 17.7 psia. Allowing for conservatism, I would recommend the use of Ll 8_psjafor natural circulation cal cuations, Additionally,'I Would estimate available core 4 bypass flow at present conditions to be 4600 to 4800 lbm/sec.I have asked John Castanes to obtain up-to-date readings on Gentille delta P's as a further check on this analysis. He has been in contact with BMCo and they feel that the transmitter accuracy is very good.ATTACHMENT PSB:jVws" ' ': , ,' , ' .'.: * ' </I I i:11 *11* .S: [j ! I TTT il * * **

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Vi.-I-.!-i7..I -I I.dc [. * -j ~ 1 I I I:, I I- l I -.1.! .I! l-1 : 1 ....-I -.I I-' !- -H-+ 1-: 1+-H -H 1h -I -I 4- - 1 I- I -- I- I .----i-,-l-I-..-t- -. -H,._a.1 a1 , ir-.i- .i ,-r It I. ;j 1 I Ub-;.;- .;;;;; .I I; .,I II -I- I., .;-- ..I .I ., I., I_ -.I...JU -! ; : : .: i :, ::1 AL ,: ' ,- .I I-;.I I;.;. i.1: 'I IiA It I. I]I~ I I : 1 1 .1 1'Il II'I'l liii... .I..LI LI.III i P.I I I1 I I.I I&I* .I I 01°*.;I , I... :_ THE BABCOCK & WILCOX COMPANY POWER-GENERATION GROUP To l.I _.J. S. TULENKO -MANAGER, FUEL ERGiMEERING BDS 663.5 ,..E R.V. DE MARS -CCTF LEADER '-WI 1/. k tn rn. rECN DITMION TASK FrnRrr Cust. " Fi le No.or Ref.SubjC. I Date CURRENT ASSESSMENT OF CORE'CONDITION 4./7/79 (1800) -7T-7:8pml1his 11tof la co-of en. enseomet old ene subject only.ATTACHED IS THE CURRENT ASSESSMENT OF THE CORE CONDITION BASED ON INFORMATION AVAILABLE AS OF 4/7/79. MOST OF THE INFORMATION USED IN FORMING THIS ASSESS-MENT, IS STILL PRELIMINARY AND REQUIRES VERIFICATION AND DOCUMENTATION. THE MOST SIGNIFICANT UNCERTAINTY IS THE TIME AND TEMPERATURE CONDITIONS PRESENT DURING THE CORE UNCOVERY.THE CORE CONDITION TASK FORCE CONSIDERS THIS A PRELIMINARY BUT REALISTIC ESTIMATE BASED ON VARIOUS SOURCES OF INFORMATION INCLUDING INPUT FROM THE EPRI TASK FORCE ON FUEL DAMAGE ASSESSMENT. AS FURTHER INFORMATION BECOMES AVAILABLETHE ASSESSMENT WILL BE UPDATED ACCORDINGLY. RKK:dww ATTACHMENT CC: D.H. ROY E.A. WOMACK C.D. MORGAN XC: CORE CONDITION TASK FORCE INTRODUCTION -THE PHYSICAL 'CONDITION OF THE CORE IS BASED ON THE FOLLOWING POSTULATED SEQUENCE OF EVENTS. THE CORE WAS UNCOVERED 'u11 FT DURING THE FIRST 15 MINUTES FOLLOWING THE SECOND PUMP TRIP. (SEE FIG. 1) THE CORE WAS THEN UNCOVEREDb FT FOR 105 MINUTES. FOR THE REMAINDER OF THE TRANSIENT, THE CORE WAS ASSUMED TO BE QUASI-COVERED TO THE POINT THAT NO SIGNIFICANT OXIDATION OCCURRED. DURING THE INITIAL UNCOVERING, THE CLADDING WOULD FAIL NEAR THE TOP OF THE ROD DUE TO STRESS RUPTURE. DEPENDING ON.THE HEATING RATES, THESE FAILURES WOULD HAVE OCCURRED BETWEEN 01200-165 0 0F. THIS MAY PRECLUDE INITIAL FAI.LURE BY EUTECTIC FORMATION BETWEEN INCONEL GRID AND ZIRCALOY RODS. CLADDING STRAINS DUE TO-fIGH-TEMPERATURE DEFORMATION-PRIORTO RUPTURE COULD APPROACH 35%. DURING THE HOLD .TIME SUBSEQUENT TO THE RUPTURE, THE CLADDING. OXIDIZED SEVERELY, FORMING ZIRCONIUM OXIDE AND RELEASING HYDROGEN GAS. THE DEGREE OF OXIDATION WILL VARY WITH THE POWER, HAVING BOTH AXIAL AND RADIAL DISTRIBUTIOiN. THE DEGREE OF OXI-DATION ALONG THE LENGTH OF A ROD COULD VARY FROM NEGLIGIBLE AT THE BOTTOM TO 100 AT THE HOTTEST REGION NEAR THE TOP OF THE .ROD.BASED ON EVALUATION AND I14TREPRETATION OF.AVAILABLE INFORMATION AS OF (4/7/79) IT IS POSTULATED THAT THE CURRENT CORE CONDITION IS: 1. FUEL ROD PRESSURE BOUNDARY APPROXIMATELY 90% OF.THE FUEL -RODS MAY HAVE PERFORATED CLADDING, ALLOWIhG RELEASE OF HELIUM AtID VOLATILE FISSION PRODUCTS.2. FUEL ROD STRUCTURAL iNTEGRITY MANY OF THE INTERIOR FUEL ASSEMBLIES MAY VIRTUALLY HAVE NO RECOGNIZABLE FUEL ROD.ARRAY BETWEEN THE UPPER END FITTING AND FIRST (TOP) INTERMEDIATE SPACER GRIDS. IN SOME ASSEMBLIES THIS CONDITION MAY EXIST TO A LESSER EXTENT AS FAR DOWN AS THE SECOND OR THIRD INTERMEDIATE GRIDS. MOST OF THE PERIPHERAL RODS AlID THE LOWER PORTION OF MOST RODS WILL BE OXIDIZED BUT NOT TO AN EXTENT TO SIGNIFICANTLY. AFFECT STRUCTURAL INTEGRITY. -I - I 3. FUEL ASSEMBLY STRUCTURE THE INTERMEDIATE INCONEL SPACER GRIDS SHOULD BE CLOSE TO THEIR ORIGINAL AXIAL POSITION. THE UPPER END GRID AND END FITTING IN MANY OF THE INTERIOR ASSEMBLIES MAY HAVE LITTLE STRUCTURAL SUPPORT. THE FIRST AND SECOND INTERMEDIATE SPACER GRIDS IN THESE INTERIOR ASSEMBLIES ARE LIKELY TO BE SUPPORTED AXIALLY FROM BELOW BY BADLY OXIDIZED GUIDE TUBES AND POSSIBLY FUEL RODS. THE REMAINING LOWER GRIDS ARE EXPECTED TO HAVE STRUCTURAL SUPPORT FROM THE DEGRADED BUT REMAINING GUIDE TUBES AND FUEL RODS.4. ZIRCALOY COMPONENT MATERIAL CONDITION THE ZIRCONIUM OXIDE (ZrO 2) PRODUCED BY THE OXIDATION OF THE ZIRCALOY COMPONENTS HAS RELATIVELY LOW DENSITY AND CAN RANGE IN FORM FROM SMALL PARTICLES OF A FEW ?41LS IN SIZE, TO IRREGULAR SHAPED FLAKES OF A FEW MILS IN THICKNESS AND UP TO A QUARTER INCH ON A SIDE, TO VIRTUALLY INTACT TUB-ULAR BUT FRAGILE SEGMENTS OF CLADDING. THE PARTICLES AND FLAKES ARE LIKELY TO BE MOBILE IN MOVING WATER. THESE PARTICLES CAN BE EXPECTED TO* LODGE IN THE UPSTREAM SIDE OF ANY FLOW RESTRICTION SUCH AS SPACER GRIDS.; GRAVITY MAY BE SUFFICIENI TO CAUSE THE LARGER ZIRCALOY AND ZrO 2 FRAGMENTS TO SETTLE OUT ON THE DOWNSTREAM OR UPPER SIDE OF SPACER GRIDS. THE QUANTITY OF ZrO 2 AND FRAGMENTED ZIRCALO'Y PRODUCED DURING THE PARTIAL CORE UNCOVERY IS LARGE. EXCEPT FOR SOME RODS IN PERIPHERAL ASSEMBLIES AND THE LOWER PORTION OF MOST RODS IN ALL ASSEMBLIES, THE TEMPERATURES PROJECTED FOR.THE ZIRCALOY FUEL RODS WAS SUFFICIENT TO CAUSE SIGNIFICANT OXIDATION. THUS, THE MOBILITY, QUANTITY AND ORIGIN OF ZrO 2 IS SUCH THAT LOCAL FLOW BLOCKAGE COULD BE EXPECTED TO OCCUR IN ALMOST ANY LOCATION IN THE CORE.HOWEVER, THE MOST EXTENSIVE FLOW BLOCKAGE COULD BE EXPECTED IN THE UPPER CENTRAL PART OF THE CORE, WHERE THE ZrO 2 PARTICLES COULD FURTHER RESTRiCT THE GENERAL FLOW RESTRICTION CAUSED BY. THE HEAVIER FUEL PARTICLES AND FUEL ROD FRAGMENTS.

5. FUEL (U0 2). CONDITION THE FUEL RELEASED FROM THE DETERIORATED CLADDING IS -VERY DENSE. THE ORIGINAL SIZE OF PELLETS IS APPROXIMATELY 3/8 INCH IN DIAMETER BY 5/8 INCH LONG. UNDER IRRADIATION, THERMAL STRESSES CAUSE THE PELLETS TO BREAK UP INTO FRAGMENTS GENERALLY RANGI NG IN SIZE FROM 1/16 INCH TO 1/4 INCH ON A SIDE. DURING A TRANSIENT'AND THE PERIOD FOLLOWING, THE FLOWING WATER AND STEAM CAN BE EXPECTED TO CAUSE SOME FUEL EROSION, WHICH WILL-PRODUCE VERY SMALL PARTICLES WHICH CAN BE SUSPENDED IN MOVING WATER IN THE-CENTER ASSEMBLIES, IT IS LIKELY THAT MOST OF FUEL HAS BEEN RELEASED-F40M-ThF ODS BETWEEN THE END FITTING AND THE ;SECOND OR THIRD INTERME'DIATE GRIDS. BASED ON THE UNDERSTANDING THAT THE FLOW IN THE CORE IS SEVERELY BLOCKED, THE FUEL FRAGMENTS HAVE SETTLED ON TO THE INTERMEDIATE GRIDS. LOCAL FLOW PERTURBATIONS CAN 1MOVE PELLET FRAGMENTS THROUGHOUT THE SYSTEM. FUEL FROM THE UPPER LEVEL MAY HAVE SETTLED DOWN THROUGH. THE TOP' INTERMEDIATE SPACER GRID TO THE SECOND LEVEL AND LOWER LEVELS TO A LESSER EXTENT. THERE IS SOME REMOTE POSSIBILITY THAT THE STRUCTURE SUPPORTING THE FIRST TWO IN4TERMEDIATE GRIDS IN THE CENTER FEW ASSEMBLIES

MAY COLLAPSE, CAUSING THE TOP 5 FT OF FUEL TO SETTLE ON THE THIRD. INTERMEDIATE GRID. THE FUEL FRAGMENTS WOULD LIKELY BE MIXED IN WITH SOME REMAINING ZIRCALOY ROD FRAGMENTS.

THE SPACE BETWEEN FRAGMENTS COULD BE FILLED WITH WATER, STEAM, ZrO 2 , OR SOME COMBINATION THEREOF.THE LARGE QUANTITY OF SMALL Zr0 PARTICLES COULD CAUSE SOME LOCALIZED FLOW*2 BLOCKAGE TO PREVENT FULL COVERAGE WITH WATER. THE PRESENCE OF SOME TUBULAR SEGMENTS COULD ALLOW LOCAL FLOW CHANNELING AND ATTENDANT "JETTING".

6. PROJECTED STABILITY OF CORE CONDITION THE POSSIBILITY OF CONTINUED STRUCTURAL DEGRADATION REQUIRES FURTHER EVALUATION.

FLOW.BLOCKAGE IS LIKELY WHICH CAN CAUSE LOCALIZED BOILING. WHEN LOCALIZED BOILING EXISTS, A FURTHER REDUCTION IN SYSTEM PRESSURE WILL INCREASE THE AREA OF BOILING-AND RAISE THE TEMPERATURE OF CLADDING IN THE-AFFECTED AREA. IF THE TEMPERATURE OF ANY ZIRCALOY COMPONENT EXCEEDS 1000 0 F., ACCELERATED OXIDATION WILL ADD TO THE GENERATION OF HYDROGEN AND CAUSE FURTHER DEGRADATION OF THE CORE STRUCTURE.

7. DISTRIBUTION OF FUEL AND ZrO IN SYSTEM .2 IT IS VERY LIKELY THAT PARTICLES OF ZrO AND UO ARE CIRCULATING THROUGHOUT 2 UY 2 THE PRIMAIRY SYSTEM AND'MAY SETTLE'OUT IN STAGNAN4T AREAS.

$1 I IAlto ~fu t i tr i: t I .X w., , ; ', ! "', i -i.1 ,* r -l8 .. I. I4. .* r., A.J I TOP 0Z fC&:-E s..* ' u,.,.- -, S -V* .t. .* 0* * *.S. g.1 ..... *. .* A.* -.1 S.* 4.... * ,. ..'v\. 9. v %.1 $.A'a.3.'.'I.I* I..4 -0* a-J 3 a.I,; a La*1..i Ii a'.3.3.a, 3.a.'* a S.a.gI I, S.3.9.I ,._I4 I _ ,.. P .*4?*.4^ :. i .I v , ,,_ V:>\?, .! ,2..-L- I- V , a' .;~~.,,.,..W}* i:.P..r)-A ,.8 6 p 's .1- " ! 9. t k t 4.a, 6 C i, .f v__ ............ ._ ... ,_ ......... ,. ,_+ .... J .w_. .............. w__ __ " *q t p 9 9 4 1'Is; _ S _ ol -t *01 .;. Ci__-I I A a.-a 9* .%..C.,.~D/5- 1 I ,- r I ;_--S *I-.-'.1 I D. O-,Jb!),-a._ _ ._ ... .. .___.. ., -A/A A ' -..I I 7...~~~~ I.. 1/ .... ...... _. .._ ....... _J .!... ,.,,, , 0 i ,., -,, t ;.7 :.! C., C ( . v --~FCI.EJ , ( CORE MELT SCENARIOS* Z&. 1 GENERAL..UNDER:THE CONDITIONS OF NATURAL CIRCULATION, THE CORE WILL BE SURROUNDED-BY COLD WATER NEAR 100 0 F. THE !POSSIBILITY OF' CORE MELT IS CONSIDERED TO BE REMOTE UNDER THESE CONDITIONS.

WITH THE CURRENT LOWl DECAY HEAT RATE AND WITH' APPROPRIATE IONITORING OF INCORE THERMOCOUPLES, THERE WILL BE SUFFICIENT EARLY WARNING SIGNALS TO PREVENT A CORE MELT SITUATION.

M6.2 CORE MATERIAL MELTING POINT THE MELTING TEMPERATURES FOR THE CRITICAL FUEL ASSEMBLY MATERIALS ARE SUMMARIZED ON THE ATTACHED TABLE. THE MATERIALS INCLUDE THE FUEL ASSEMBLY STRUCTURAL MATERIALS (END FITTINGS, GRIDS, HOLDDOWN SPRING, AND GUIDE TUBES) AND PRIMARY FUEL ROD MATERIALS (PELLETS, CLADDING, AND END CAPS). IN ADDITION, THE MELTING TEMPERATURE OF THE CHRONEL ALUMEL THERMOCOUPLES IS ALSO INCLUDED.. ..E.. 3 .ASSESSMENT OF ORIGINAL FUEL DAMAGE CONDITIONS DURING THE INITIAL ACCIDENT, IHEN THE CORE WAS PARTIALLY UNCOVERED, THE THERLALL CONDITIONS WERE VERY SEVERE. HOWEVER, THERE ARE INDICATIONS THAT THE CORE DID NOT UNDERGO MELTING.THE DECAY HEAT.RATE WAS IN EXCESS OF 25 1 .ANY WATER NEAR THE CORE WAS NEAR SATURATION TEMPERATURE OR %500-650 F. UNDER* THESE CONDITIONS THE FUEL ROD CLADDING REACHED 2000F OR HIGHER AND OXIDIZED SEVERELY TO PRODUCE HYDROGEN. HOWTEVER, CONTINUED OPERATION OF THE INCORE CHROMEL ALUvEL THERMOCOUPLES, WHl1ICH HAVE A MELTING POINT NEAR 2500F, INDICATE THAT :THE STEAM TEMPErUATUPE INSIDE.1OF THE INSTRUXENT. TUBE WAS LESS THAN 2500F. THIS IS APPROXIMATELY 250 OF FROM THE MELTING POINT: OF. UO THE CENTER-2'LINE TENPERATURE OF U0O PELLET FRAGMENTS IS ESTI'MATED TO BE NO MORE THAN 1QOF HIGHER THAN THE STEAM; THUS; SHOWING A LARGE MARGIN TO U0O MELTING. LOCAL HOTSPOTS 'FUEL MAY HAVE BEEN HIGHER* FR HOT ERIDSBU TE ESTIMATED 20FMARGIN TO MELTING WAS SUFFICIE1NT TO PRECLUDE MELIING.*THE RADOCHENISTRY ANALYSIS OF B 140 AND OTHER ISOTOPES IN COOLANT SAMPLE TAKEN A DAY AFTER THE-ACCIDENT DID NOT INDICATE THAT U0 2 ELTI1NG HAD OCCURRED.EARLY WARNING SIGNALS DURING THE TRANSITION TO NATURAL CIRCULATION, THE INCORE TRERMO-1*COUPLES W7ILL BE MONITORED. THE TEMPERATURES ON THESE THERMOCOUPLES' HAVE A NORMAL READOUT RANGE OF UP TO 900F. SINCE SOME LOCALIZED BOILING IS EXPECTED, A FEW OF THE THERMOCOUPLES CAN BE EXPECTED TO READ HIGHER THAN SATURATION TEMPERATURE. HOWEVER, BECAUSE OF THE SLOW HEAT UP OF THE OVERALL SYSTEM, THE MAJORITY OF THERMO-COUPLES ,AS A GROUP, CAN BE USED TO MONITOR THE. BULk COOLANT BULK COOLANT TEMPERATURE ATLTHE TOP OF THE CORE. LTEQPERATURES APPROACHING SATURATION TEMPERATURE WOULD BE AN EARLY INDICATOR THAT LOCALIZED BOILING WAS SPREADING AND THAT CORRECTIVE ACTION SHOULD BE TAKEN.LARGE AROAS F MBOILING.ARE UNDESIRABLE SINCE THEY LEAD TO HIGH OSTEAM TEMPERATURES. WHEN THE STEAM EXCEEDS 1000F, THE ZIRCALOY (1 -COMPONENT WILL BEGIN TO OXIDE AND PRODUCE HYDROGEN. AT HIGHER TEMPERATURES THE RATE.OF HYDROGEN PRODUCTION WILL INCREASE.HOWEVER, AS INDICATED IN -%5.3, EVEN WITH THE HIGH STEAM TEMPER-ATURES PRODUCED DURING THE INITIAL CORE UNCOVERY, THE UO 2 DID NOT MELT. THUS, THE INCORE THER1IOCOUPLES CAN PROVIDE EARLY-WARNING SIGNALS SUCH THAT CORRECTIVE ACTION CAN BE TAKEN TO PREVENT A CORE IELT SITUATION. i!II Material U0 2 Zr-4 Inconel X-750 Inconel 718 CF3MSS (and fitting)* Zr02.Chromel Alumel* .TABLE Zi4.'Melting Points of Core Xaterials Melting Point F 5081 (1)3353 (2)2570 (3)2323 (4)2550 (5)*5010 (6)2500 (7).Oxidation of Zircaloy is assumed to. initiate at 1000 F; : 1. Hausner, H., "Deteraination of the Melting Point of Uranium Dioxide",_ Journal of Nuclear Materials, Vol. 15,1965.2. Hammer, ri. R., In-1093, Sept., 1967.3. Huntington Alloys Technical Bulletin, Inconel Alloy X-750, (1970).4. Huntington Alloys Technical Bulletin, Inconel Alloy 718, (1968).5. Materials Engineering,. Materials Selector 77.6. Lynch, J. F., et al, Engineering Properties of Selected Ceramic Materials, (1966).7. Weast, R. C., CRC Handbook of Chemistry and Physics, 48th Ed, (1968).}}

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