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Forwards Summary of Input Assumptions Used in NPSH Calculations,Positive Suction Head Calculation,Evaluation of AEC 701102 Safety Guide 1 & Possible Resolutions If AEC Enforces Safety Guide 1
	ML19308D692
Person / Time
Site:	Crystal River
Issue date:	12/13/1971
From:	Vaughn R; GILBERT/COMMONWEALTH, INC. (FORMERLY GILBERT ASSOCIAT
To:	Obrien W; Office of Nuclear Reactor Regulation
Shared Package
ML19308D691	List: ML19308D690; ML19308D692; ML19308D694;
References
	14374, NUDOCS 8003120823
	Download: ML19308D692 (5)
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' introduction .. Da it. t: ::.. ~ r n :... ;. *,.'t : -.. ; ~ The EC issued Safoty Guide IIo. I cn !!ase=ber 2,1970. This in a (cuido) sug3ceted cotbod of cniculating the Het Positiva Suetics I! cad (11?:7.1)' for the cr.nr;7.:ncy coro coolina cad ccntai== cat heat rcroval syste:2 pu=os (son-attah ).- k's e-ly call rhme As tha Ecactor Du11dina Upray (R:1S) t and Decay llont Reacnral (011) peops. Uwe asst:sa the debatable nasu=ptions in Safoty Guida r.o. I are correct it is possihlo that the present n.3.S. and D.!!. ptz:7s ecy not hava sufficient EPSIl to sarisiy tha EC Salaty Guida requirments. Theco nyatnes vera desi,v:ad and the current pt==ps purchased prior to tim is:n.u:nce of tua safety cuido IIo.1. Ua now havs 1. Pericued these systen denicas aconnin;g the requirrnmM act forth .. in Saicty Guida Iso. I hold. 2. Vo also performed prolininsry calculatic is to deter:sino t.hother or not tha '.'. 3.S. c..d cita i).11. re cial n rate : p: na h.:d e diicient li?.TJ.Trailable to catisfy clu S ciety Guido require?. ants. ha results of the:o prelicinary calculationa arn chova in the attachnent to this Iceter. e Calculatio-.s Perfz. sed l Several Incs-of-Ccolant-Accident (14CA) rceircul:t$1on conditions havu been' n.a.217:cd. 02 these the taret caso is d:a >oct LLCA rceirculatica ccads.tien calculated in GPS 11 Calculacions Soc.:ica III (pa,cu 23 to 32) of tho atta ch-8003120 f j ~~ C"f ~ '" ....j w

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P1arida Powee Corporation Decembee 13, 1971 ' FPC 5043 ' Pass 2 Ecst. %is vorst possible caca assumes a bres!: in tha reactor ecolant piping at the top of the acca:n nencrator (clevation 177'3"). ha primar7 icop cco. poets vill retain, and vill not loose, their vster with a break. at this cicvation. 1:sdcr this condition less water is available to ficod tha reactor building floor.resulting in a lower untar level within tha reactor, building su=p. - ; ~. :.~1.- Calculation Results fo-Uorst case NPSH Calculatio.s Section III --r- ~ A deficiency of.62 = (23.34 ft - 22.72 ft) ft H,io in U?SH availabla is real-i=ed under tha above recirculation condition (scE Tabla V ite:a 2 of attach-ment). .s. Duo to the cancervativa assu=ptions and the fact that en estimated (probably conservative) UPSU value of 13'0" Ii20 is used as a requirc=cnt for the re-actor building sprn.' pumps (RBSP-33)(refer to I?orthington pe=p curva E-135235 ^ dated 5/9/67) the roti=sted LPSH therefore calculated and given in ?sbles I throus V, inclusive, is c=pected to be larger than the actunt UPS11 required (appro=inately 11'0" H O). lia believe the I:FSH deficiency of.62 ft H O vill. 2 2 then be core than cocpensated for, and va probably vill ha 01. Calculation nesults - N?sn Calculations Sectien I De calculations perfor=ed in this ocction of' the attachment ara nada with the ascu=ption that no ascrqcacy cooling water is retained by the pri=ary cysten cocponents. This allous a larger vole =a of H,,0 ta flooa the reactor building su=p thus a higher cicvation of water relacIvo to tha reactor build-ing cpray and the decay heat' removal pu=ps cuction ccatcrlina. Under these. condition?, adcquata 1:PS3 is availablo. Results of calculacions ara chos.n ~ on Tablo I of the attachment. Calculation Renults - r*Ft Calculattens Section II na calculations perfor=cd in this cection are cada with the aanu:ption that a breati occurred at the reactor vessel ec::les. tiith this nas_- pcien, ade-quato NPSU is available to tho reactor buildin3 apray and tha decay heat recoval pucps. Ecsults of calculations are shotnt on Tablo 17 of tha attach-ccat. Conclusion The only danner of the ccaracncy core cooling (D11) aol contair:2 cat heat ro'- Safety Guida 2:o. I requirements, if this turns cut to bo' t;ta actual cass,11. _ _, coval cyaten G.M) pt:2p3 not havina suf ficient *i/S3 availabia to satisiy a ,,,..+r-

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sotaly dependent on the allcr.rsbie reactor building vatar Icvel, (i.e. static unter heads.dovo the Pu=p suction pipe centarline), for titich thm AEC (Safety Guida 1 essu:::ptions) vill allev credit to ba ta'.:en in calculating the availablo 1.7511. In essence, with AEC Safety Guida 1 assu=ptions, any arbitrarily imposed water IcVet other than the calculated =inicu:2 post LCf.A vater level in Sec-tion III of the attachnent could seriously inpair the ability of tha pumps to perform their safety fu=ction. ~ Arbitrarily i= posed H O level are discussed and evalunted in the following 2 attach:cen% If there are.'ny questions, please advisa. .-s..,.- ............... Very truly y.ours, ~

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\\ ( ab,i. ',, ' SIDDIARY OF INPUT ASSUMPIIO;iS USED, RESULTS OF NET POSITIVE' SUCTION !!EAD CALCULATION, BRIEF EVALUATION OF AEC SAFETY GUIDE NO. 1 DATED NOVEMBER 2, 1970, AND POSSICLE RESOLUTIO!!S IF AEC INFORCES SAFETY CUIDE.MO. 1 The NPSil availabic to the decay heat and reactor building spray pump.s during ~ post LOCA recirculation phase has been calculated based on " Issued for Con ~struction" piping drawings and the follouing assumptions: Pipe and fitting losses calculated using the information in Crane n. Technical Paper No. 410. b. Total required flow in a single string (i.e. consisting of one decay heat and one reactor building spray pump which have a common suction - line from the reactor building sump) which is 4500 gpm or 3000 gpm to the decay heat pump and 1500 gpm to the reactor buil. ding spray pump. ,c. Sump water temperatur6 equals 226 F (peak value from Figure 14-61 FSAR). d. Reactor building pressure equals saturation pressure at a temperature O of 226 F. Water Icvc1 in the reactor building is at 99.82 ft. This water Icyc1 c. in the reactor building is derived from preliminary calculations in which all sater capabic of flooding the reactor building and all dis-placements of water such as components, sumps, tunnels, ete have been analyzed. This wat'cr Icyc1 will be referred to as the post loss-of-coolant accident (LOCA) water Icyc1 throughouc t.his HPSII revicu. The NPSH's available to cach pump, as calculated on the above basis, are com-pared belou uith the NPSil required by cach pump at the f, lows indicated above. ' Rcquired NPSl! is from curves furnished by the pump manufacturer. The requir'ed NPSil in feet of 1120 is an approximate figure, it is expected that the actual I pump performance of NPS}l required vill be somewhat 1 css as compared. to thus -- w 'l .1 w n h i

+ g - - - - - - - - - -.7 3,,. - + figure shown below. Calculated NPSif Flow Availabic ft U O NPSH 2 Ra.te ' El. Head.+ Vel. Required Pump gym. Head f t M O_ 2 R.B. Spray Pump A (BSP-3A) 1500 13.25' 13.00' R.B. Spray Pump B (BSP-3B) 1500 12.38' 13.00' _ 1 D.H. Pump 3B (DHP-3B) 3000 16.46' 13.50' 114 D.H. Pump 3A (DHP-3A) 3000 Similar 13.50' The above method of calculating available NPSH is in agre'ement with AEC Safety

Cuide No.1 in that no credit it taken for blowdown pressure in the reactor building.

The Safety Guide states that "no increase fri containment pressure from that present prior to the postulated loss-of-coolant accident' may be assumed. C This position can readily be applied when c'alculating NPSH available with a sump water at a temperature less than 212 F. This po'sition [s* unrealistic when the sump water is above 212 F in that it would 'mean the entire large inventory of water in the reactor building would boil away without increasing the pressure in the reactor building. It is more realistic to ass'ume reactor building pressure will achieve equilibrium with the sump water at t a saturation pressure of the-sump water. Any abrupt loss of saturation pressure in the Contaitunent Vessel would result in flashing of the sump water at temperatures above 212 F and Jm-0 mediate equilibrium of the reactor building pressure with the cump water at saturation pressure.vould be restored. There'are two potential po,t LOCA vater elevations in the reactor building which AEC could impose other than the calculated minimum post LOCA water elevation of i 99.82 ft, currently being used. These are the top of the reactor building ' sump (elevation 95'-0") or the centerline of the outlet pipes from the reactor build-l l ing sump (elevation.86'-3") y+ we, h. r.:#.(W$tp'I4"'4 2"rJ M '~M-A ~ .y - .m __g.7 ogw

'u 4 Thn effrct o[ thesa louir cicvations of, the water icvel, uithin the reactor building on the available NkSII vs required NPS11 are i:ompared on Tables II arid ^ III. Assuming a reactor building water clevation of 95'-0" (top of reactor-building / sump) in reference to the reactor 'ouilding,sp* ray pump suction centerline eleva-tion of'78.5", the total elevation head available plus the existing velocity head at design f' low would equal approximately 17.9 f t. As noted in Table II the total required head.to the pump is 22.47 ft.

Thus, by calculating the available NPSH to the reactor building spray pump (R.B.S.P.-

3A) with a water elevation of 95.0 ft, in the reactor building sump a deficiency in net positive suction head available of 4.57 ft is realized. I Therefore, in conclusion, any arbitrarily imposed unter elevation other than the currently calculated post LOCA water elevation would ' severely effect the net available NPSH. In order to remedy the inadequate available NPSH, due to water elevations lower than POST LOCA levels, it appears that major design changes and/or modifications to the present design would be necessary. ~ In the event that the elevation in the reactor building sump is to be a maximun of 95.0 ft., the following alternates could possibly be used in resolving the 1 inadequate NPSH. Exchange present pumps and purchase new pumps with an NPSH requirement a. of approximately 8.0 ft. ~ b. Relocate pumps from the present design location to a lower elevation, in reference to the Rx building sump clevation. Design an independent and mechanical means of increasing the net avail-c. abic NPSl!. e e 94 3

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7.f ~ l 9. NPSH CALCUId.TIONS SYSTEM 1 1 1 i l Review Calculations and Results regarding the' height of the II 0 in the Reactor 2 Building. The follouing preliminary calculations were perfor.med with the assump- .q tion that all the H O available 'will flood the reactor building floor 2 following a 100A. Results by calculating the available NPS11 in. this manner indicates the highest possible H O level within tI reactor 2 building. ) C O ee* e G O e e e D G 9 9 O S e O l l g .9 w.,.. 4 4 e

Tatc1 110 in Rector Systcm Under Nermal Operation (hot) ~ 9 Reactor Vessel 4,058 ft3 Steam Generator (2,030 ft ) each (2) 4,060 ft3 Pres'urizer s 800 ft3 Reactor I'nlet Piping 3 1,085 ft ' Reactor Outlet Piping 979 ft3 Pressurizer Surge Pipe 3 20.ft Pressurizer Spray Pipe 2 ft3 Reactor Pump Casing (98)'(4) 392 ft3 11,396 ft3 Total Water Available/ Sources Outside the Normal Operating' Primary System Makeup Tank 600 ft3 ~ Core Plooding Tanks (2 tanks) 1,880 ft3 Borated Water Storage Tank 50,800 ft3 Sodium Thiosulf, ate Tank 1,600 ft3 Sodium Hydroxide Tank 1,500 ft3 56,380 ft3 Total Cubic Feet of Water available and capabic of Flooding Reactor Building floor 56,380 ft3 _11,396 ft3 67,776 ft3 e b S

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e' 9 The following components will help in lowering the height of H O in the 2 Reactor Building since they displace or are filled with water: Instrumentation Tunnci V 4' x 4' x 40' 650 ft3 R.C. Drain Tank V (4')2 (22') 1,101 Building Sump Volume V 10' x 10' x 9' 900 2,641. It3 _Hence the net volume available and capable of flooding the reactor building - floor-V 67,776 ft3 - 2,641 ft3 = 65,135 ft3 e s e 9 O e k a e g e e e e e N 4 g .i 4

-e Tot,1 Area displaced by Components within the Reactor Huildinn ~ Steam Generator (2) A = 5.14 (6)2 = 113 226 Pc2 Pressurize.r = 3.14 (4)2 = 50 50 Primary Shielding A'= 3.14 (16.5)2 = 855 855 Main Coolant Pumps A = 3.14 (3.75)2 = 44.2 44.2 Main Coolant Pump Pipes (Vertical 4) A = 3.14 (1.5)2 = 7.07 28 Misc. Shielding .A = 13.x 5 65 = ~ 2 2 = 3.14 (14 -11.5 ) = 201 = 45 x 5 (2) 400 = = 62 x 2.5 155 = = 10 x 9 90 = = 75 x 1.5 113 = 2,410 ft2 G S O 7 g -, - +

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y, y, g a c,1culnted Voke <te. of teedge Surreunding Reector Building Baso 9.0 - 5.6' = 3.4' c tancC.= x c x .'. a + b "' " a +c a a+b; tanot ^ = a a b (3.4)(1.5) [ 5.1 X " .567' = 3.4 + 5.6 9 = 1/2 bh = 1/2(.567')(3.4') Area triangle A 2 .96,4 ft = Area triangle B 1/2 bh = 1/2 6.75 ft2 ~ = = Cire. Rx Bldg. = Y d 3.14 (131') = 412 ft = e Total Volume displaced by surrounding wedge Arca A A - Area A B ~ 5.804 ' = ~ Volume displaced = 5.804 ft2 3 x 41,2 ft = 2389 fit e .e e D e O e l 1 i ~ 9 t' I

(. ~. (. Total Volume it O followino: LOCA vs Height in Reactor Building 2 Total Volu::te of H O uhich is vapor = 5'/. of 65,135 f t3 = 3260 ft3 2 ~65,135 - 3260 = 61,875 fe 3 Total H O in Rx Bldg. following LOCA' 61,875 ft 2 2 13,460 ft Total Area of Rx Bldg. floor = ^ 3 To'tal volume displaced by wedge = 2389 ft 2 ~ Total area displaced by components =.2410 ft V h I = = (61,875 ft3 + 2389 f t ) 3 h = 13;460'ft2 - 2410 ft2 c 3 64,264 ft ~ .h 5.82 ft = = 2 11,050 ft 9 5. D o .m s Tb o.B D. e a e e e e e 10

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s Available NPSH vs Required NPSH with Total H O Available 2 Hence; maximum H O level in the Reactor Building is establi::hed 2 (All Reactor Bn41Mng components empty except the reactor' vessel) Based on BSP-3A & 3B and m P-3A & 3B Operating at 1500 gpm and 3000 gpm, respectively Total Elevation Potential Elev. Head Available ' Total Head Required Head Available on this Basis to Plans - Ft M 1. From maximum post IDCA water cicvation to centerline of BSP-3!L Pump Suction 100.82' l.'40' Vel. Ed. 9.47' Ed'. Loss to 78 5' W 22 32'Elev. Ed 13.00'N MH Req'd. 23 72' 22.47' p u. e i 2. From ma h. post LOCA water mH elevation to centerline of 1.40' Vel. Ed 10 34' Ed. Loss - VSP-3B Punp suction 100.82' 22.32',Elev. Hd 13.00' NFSH Req'd. to 78 5' 23.72' 2'3 34'- 3 From maximum post LOCA water i elevation to centerline of 8 MF-3D Fump Suction 100.62' 2 30' vel. Ed 9 16' Ed. loso d [fLQj - to 77 5' 23 32' Elev. Ed _13 50' Ese)? 'g' Q: 25.63.' . 21.66' 4. M P-3A Similar u. a., mM Ess ?sB e a I 1:

. h.. 12)'1 TABLE II Available NPSH vs Required NPSH i J This table shows the available NPSH vs the required NPSH in the event AEC does impose a mnviam water level of 95' (clev.) which is the top of the reactor building sump Total Elevation . Potential Elev. Head Available .Tctal Head Required Head Available on this Basis to Pumps Ft 1. From Top of. reactor building (R.B.) Sump to Centerline of BSP-3A Pump 1.4' Vol. Ed. 9.47' Ha. Ioss Suction 95' - 78 5' 16.5' Elev. Hd. 13.00' NPSH Required. '17 9' 22.47' 2. From Top _ of reactor building (R.B.) Sump to Centerline of BSP-3B Pump 1.4' Vol. Ed. 10 34' Nd. Loss p Suction 99' - 78 5' 16.5' Elev. Ed. 13.00' HPSH Req'd. m 17.9' o 23 34' 3. From Top of reactor building (R.B.) Sump to Centerline of IMP-33 Pump 2 3' Vol. Ed. 8.16' Ed. Loss Suction 95' - 77.s' 17 5' Elev. Ed. 13 50 19.8' 21.66' 6

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/, 12/171d gg m .~ d Availabic NFSH vs Required NPSH This table shows the available UT3H vs the required NFSK in the event AEC does i..: pose a maximum clicwable water level of 86.25' '(Centerli'ne of reactor building sumn outlet piping) ~ Total' Elevation . Potential Eley. Head Available Total Head Required Head Available on this Basis to Pumns 1. From Centerline R.B. Sump. m outlet to centerline of BS'P-3A 1.k' Vol. Hd. 9 47' Ed. Loss Pump Suetions 86.25' - 78 5' 7.75' Elev. Ed. 13 00' NFSH Req'd. 9 15' 22.47' 2. From Centerline R.B. Sump outlet to centerline of BSP-3B 1.4' vel. Ed. 10 34' Ed. Loss

Pur.p Suction 86.25' - 78.5'

'7.75' Elev. Ed. 13.00' NPSH Req'd. 9 15' 23 34' 3 @, d From. Centerline R.B.' Sump outlet to encterline of DHP-3A 1.4' Vel. Ed. 8.16' Ed. Loss g Pug Suction 66.25' ~ 77 5' 8.75' 13 50' UPSH Req'd. 10.15' 21.66' ^ M y @ Kote: For AEC to deter.ine elevction level 66.25 feet cs the :r.c:d:mm cllc,.able vctor level to be u:ed in calculatig the evcilabic E2SH vould be unreclistic,.in that at this elevatio:t l g level there is an incufficient volur.: of veter available to the pur.ps. a y i e 1 i j' n,

.,s....- ( ( NPSil CAI.CULATIONS SECTION II Results of a reactor coolant break at elevation,128'-0". ~ The following preliminary NPS:t and 110 height within the reactor building 2 calculation was perforced with the assumption that a primary piping break occurred at the reactor vessel nozzles. Calculating a breaI: at this cicva-tion vould mean that a minimum of 1I 0 uould be retained within the l>rimary 2 System components hence, the post LOCA water icvel would be at its maximua in the reactor building. y p g ee 8= t l e 1 i l \\ 1 )l E I T3liR ~ g e? \\ Sa

j Postulating a reactof coolant break i.t the reactor coolant inlet and/or outlet pip?ng to the reactor vessel at elevation 128'-0". _ Volume of H O retained within the reactor building components 4 Steam Generators would retain.- 998 f t3 (2) = 1996 ft3 ~ Reactor inlet piping wou1d retain 979 ft3 Reactor vessel would retain 3080 ft3 Total retained H O 6055 ft3 2 Total H90 available (storage) Makeup tank 60.0 ft3 Core flooding tanks (2 tanks) ^1880 ft3 Borated water storage tank 50800 ft3 Sodium thiosulfate tank 1600 ft3 Sodium hydroxide tank 1500 ft3-56380 ft3 Total cubic feet of water available and capable of flooding r'eactor building floor 61721 ft3 2 e e 9 O 't 4 e f 9 l e 15 2

\\ \\ The following ca.nponents will help in lowering the height of H O in the reactor 2 building, since they displace or are filled with water. Instru :entation Turmel 640 ft3 P..C. Drain Tank V2 (4 )2(22') 1101 ft3 Building sump volume V 10' x 10' x 3' 3 900 ft 3 264'1 ft C s. WIH PIIB La s .u 9 16

TOIAI. AREA DISPLACED BY COMPONh2 TIS Sec.~.:a Generator (2) ~ A= (6)2,. 113 226 ft3 ~ Pressurizer A= (4)2 50 = 50 Primary Shielding A= (16.5)2 = 855 855 Main Coolant Pumps A= (3.75)2 44.2 44.2 = Main Coolant Pump Pipes (Vertical 4) 2 A = < (1.5) 7.07 = 28 ~ Misc. Shiciding A 13 x,5- = 65 = (142 2 x 11.5 ) = 201 = 45 n 5(2) = 400 = 52 x 2.5 = 155 = 10 x 9 = 90 = 75'x 1.5 = 113 1074 = 2410 ft2 TOTAL AREA CF REACTOR BUILDING FLOOR r2 3'14 (65.5')2 A = = 13,460 sq ft = \\ G t

m t' H O available The height of H O in reactor building = 8 28 2 2 Total area of R.B. floor 3 y 61.721 ft - 2641 ft3 .h. 59,080 h.A 13.460 ft2 - 2410 fc2 11,050

57. of Vol. = Vapor = 2,954 ft3, 59,080 - 2,954 56,126

~ ~ 11,050 11,050 / Total Height in Reactor Building = 95 + 5.08 ft = 100.08 ft. O b l l e m:0 ~ 3' f h' 12 a s IId.m ou 6 18 .- l \\

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TorAL VOLIME DISPLACED BY WEDGE 9.0' - 5.08' = 3.92' ~ cance = c e,na = x c x ae = ..x = ,a + b ,a+b a-a a+b (3.92)(1.5) 5.88-x - = = .555' = 3.92 + 5.08 9 s Area d A 1/2 bh = 1/2 (.655')(3.42) 1.28 ft2 = = 1/2 = 1/2 ( * }(' } = 6.75 ft2 Area-d B = t . Circumference Rx buildin'g = if d 3.14.(131') I12 ft = = t Total volume displaced by surrounding wedge Area triangle A - Area trinagle B = 5.47 ft2 Volume displaced = 5.47 sq ft x 412 ft = 2260 ft3' D e s WI5Y EU o, m& IJl,Anr

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.-'y TOTAL H O AVAILABLE VS 'llEIGHT IN Rx BUILDING (incidding volume displace by 2 wedge) 56,126.+ 2260'= 58,386 ~ 58,386 h 5.28 ft = = 050 .~ TOTAL HEIGHT H O IN Rx BUILDING ~ 2 95 ft floor elevation -F 5.28 f t (H O) = 100.28 fe eley '# 2 ~ O O 9 e 4 0 O e G e e S 4 9 e e O O 3M3RR 21 b

IW. 12/1'/.72 TABLE IV Availabic NPSH vs Required HPSH Postulated break at El.128'-0" s Based on Building Spray Pump BSP-3A & 3B ond Decay Heat Pump M P-3A & 3B Operating at 1500 gpm and 3000 gum Respectively Total Elevation Potential Elevation Head Available ~ Total Head Required Head Available on this Basis - ft _to Pudps - ft ~ 1. From post LOCA wate r elevation ' to centerline of BSP-3A 1.40' Vcl. Hd. 9.47' Hd. Loss Pump Suction 100.28 to 78 5' 21.78' El. Ed. 13 00' NFSH Req'd'. 23 18' 22.47' 2. From post LOCA water elevation g to centerline of BSP-3B 1.40' Vel. Ed. 10 34' Hd. Loss Ptm!p Suction 100.28 to 78.5' 21.78' E'_. Ed. 13 00' HPSH Req'd. '23.18' 23 34.' g 3 From post LOCA water elevation .~ to centerline of MP-3B - 2 30' Vel. Ed. 8.16' Ed. Loss q'. trull yu.:p Suction 100.28 to 77 5' 22.78' Ele. Ed.. _13 50' NPSH Req'd. 3)CJ m jq as.08' a1.ss' c]g 4. mP-3A similar

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g g '., 1 9 , NPSH CALCULATIONS SECTION III Results of Postulating a Reactor Coclant Piping Break at Elevat.lon 1N7'3" The following preliminary NPSH and H O level within the reactor building cal-2 culations were performed with the assumption that a piping break occurred at "the top of one or both steam generators. (elevation 177'3"). A" break in the ~ piping at this elevation would mean that all the primary system er aponents s would retain a maximum ambient of H 0. Hence,' the post LOCA water level would 2 be at its minimum within the reactor building. e O i 9 s 4 e b G g e e e 9 9 9 9 1m p*. y1 3 g 9 23

t r \\ ( e Results of Postulating a Reactor Coolant Piping Break at the top of one of the Steem Generators - Elevacion 177'-3" Total 110 ncactor Systen under llornal Operations which tEuld bn retained. 2 Steam Generator (2,030 f t?) each (2) 4,060 ft3 t Reactor Vessel 4,053 ft3 Pressurizer-800 ft3 Recctor Inlet ' Piping 1,085 ft3 Reactor Outlet' Piping' 979 ft3 Pressurizer Surge Pipe 20 ft3 Pressurizer Spray Pipe 2 ft3 Reactor Pump Casing (98)(4) 392 ft3 ~ '11,396 ft3 1 s e e e MIRV "HHF ~ ]o 'l?ligfL, 24

R. "..:, v \\ [ TOTAL WATER LVAILABLE e ~ Makeup Tank 3' 600 ft s. Core Floo, ding Tanks (2)' 1,880 ft'3 Borated Water. Storage Tank 50,800 ft3 Sodium Thiosulfate Tank 1,600 ft 3 Sodium Hydroxide Tank 1,500 ft3 g..~' 56,380 Total Cubic Feet H O Available and capable. of 2 f1poding Reactor Building Floor 56,380 t. t c m - ~ g 25

/ .t \\ The follouink components will lower the height of 11,0 in the Rencto.- building 3 since they displace or are filled with H O_. 2 Instrumentation Tunnel 640 ft3 R.C. Drain Tank 1101 ft3 Building Sump .900 ft3 . 2641. It3 t 4 e I VIH. IIB n%.., Jy ngLu. r a. t 26

m, m -r . s _ TOTAL AK% DISPIACED BI COMIONE.*TS ~ Steam Generator (2) 226 ft;E Pressurizer 2 50 fb ~ Primary Shiciding 883n2 Main Coolant Pumps hh.2 ft2 Main coolant Pump Pipes (vertical 4) 28 2 Misc. Shielding 1,074 ft2 2,410 ft2 'JOTAL APEA OF PICTOR BUTrnING FICOR 2 = 3.14 (65 5')2 = 13,400 sq. ft A= x s O O b L% S e =M 27

u .y ~ e,. e, Total volus.2 er H O Availablo ) Dight cf H O in reacto'r building = 100aj arc (L Ol' It 52 2 .. t100r h=E h = 56,380 ft3 - 26' 1 ft3 53,739 n 3 Y 4 13,460 ftd - 2410 ft2 " 11,o50 ets ~ ~ 57, or volwt.c (H O) = vapor = 2687 ft3 2 h = 53,739 ft3 - 2687 ft3,51,052 ft3,, 11,050 fte 11,o5o pg2 1.62 ft 5 TOML H O IEIGHT IN RMCTOR BUILDDiG = 95' + 1.62I = 99.62.' 2 6 9 9 e e e b e S 6 s 4 6 e t O 4 e e O e e l e gg8L '23 e

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=v, , v.. u Total HoO Availabic vs hei,a.ht iip 0 in Reactor B R11n,3 (including volut:e displaced by veda;c) 51,052 ft3 + 2120 ft3 = 53,172 ft3 h=- = 4.82 ft O Total height }Ip,0 in Rx Bldg. = 95' + 4.82' = 99 82' e 4 e O e 4 0 e 6 9 8 e e o e r O + g O e 0 4 e 0 0 WlH '"11B

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TABLB V ?-C Available NPSH vs Required HPSH Postulated Break at Elevation 177'-3" Total M evation ~ Head Available Total Head Bcquired Potential Head Elev. Available on this Basis ft to Pumps ft 1. From top of Reactor Building (R.B.) Sump to centerline of Building Spray Pump 1.4 Vel. Ed. ' 9.47' Rd. Loss (BSP-3A) Suction 99.82,78 5'. 21 32 El. Ed. v - .^ 7: <13 00' IESH Req'd. 22 72' 22.47' 2. From top of Reactor Building (R.B.) Sugip to centerline of Building Spray Pump 1.4 Vel. Ed. ec 10 34' ud. Loss BSP-33) Suction 99 82 78.5

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.'13 00 NPSH Req'd. ~" u 22.72.' 23 34' 3 From top of Reactor Building (R.B.) Sump to centerline of Decay Heat Pump (IEP-3B) 2 3' Vol. Hd. .... 8.16'-Ed. Loss Suction 99 82 - 77 5'. 22.32' El. Ed/ g fp. 13.50' NFSH Req'd. ~ @Q 24.62' 21.66' DHP-3A Similar ,j,.1 m S Tote: Thic table chcus the availcble NPSH vc Required KISS when the post toc H ' level is at its 2O g .em1=m. g @D A noted in the cbove Tcble V No. 2, the building spray pump (BSF-33) has insufficient availabic p NPSH. This is a result of a required estincted NSSH of 13 00 ft H 0 'Et is believed that cetual g required MISH of thcae pumps (performe.nco test data) will comewhat less than the estincted ficere 2 chown,on this tc'olo. Thus, providing sufficient available not positive onetica h d, 9 I e e

S.;.J ( ( S.WETY GUIDS 1 - IEf POSITb!E SUCfIO?i HEAD FOR ) EfER3Ei!CY CORE CCOLI!iG /SD COiT/CDDDI!T HEAT RaiOVAL SYSTEi_IibuI I WIBVFill A. Introdu_ction Proposed General Desi n Criterion 41 requires that the emercency cooling 6 - and containment heat removal cystems be capabie of accomplinhing their required safety functions assuming partial loss of inctalled capacity. In current designs the ability to accomplish these safety liinctions reliably depends in part on the proper performance of systcIt pmaps which, in turn, depends on tne conditions under which the yumas nu'afdperate. One of these conditions is suction pressure. This gttido describen a. suitab3.e relationship between increases in containment prensure caused by postulated loss of coolant accidents and ~the not ponitive atectio'n head (NPSH) of emergency core cooling and containnant heat re:.= val systeat pwnps which may be used to implement General Design. Critorion 41. B. Discussion A significant consideration related to emergency core cooling anct containment heat: removal systems is the potential for deg2adell pinap performance which could be caused by a nimiber of factors, including inadequate NPSH. If the HFSH available to a ptcip is not sufficient, cavitation of the pumped fluid can occur. Y is cavitation na3 reduce si nifice.ntly the capability of the synten to a::co.r.plish ite.Mety E i\\tnctions. It is important that the proper perfornance of emergency core coalit g and containment heat removal systenn be independent of calculated increa.kes in containment pressure cauced by postulated loss of coolont accidents in order to accure reliabic operation under a variety of p..icnible accider.t-e n

/ ( \\ . _ e,.,. (- iar... conditions. For example, if proper operation of the emergency core ~ cooling system depends upon maintaining the containment pressure above a specified minimum amount, then too lou an internal pressure (resulting from impaired containment integrity or oper'ation of the cont.ainment heat removal systems at too high a rate) could significan517 affect the' ability of.this system to accomplish its safety functionn by causing pump cavitation. In addition, the deliberate continuation of a. 7.i.gh con-tainment pressure to maintain an' adequate pump HPSR \\tou1d r sult in greater leakage of fission products from the containment end. higher po-tential offsite doses under accident conditions than would otherwise result, Changes in NPSH for emetgency core cooling and containment heat reuoval system pumps caused by increases in temperature of the puraped fluid. under loss of coolant accident conditions can be accor.vnodated yrithoutylianeg c on the calculated increase in containment pressure. Adequate NPSE can be assured by locating pumps at suitable elevations uith respect to the storage volumes connected to their suction sides, by using cultistage or booster pumps, by.a combination of these nethods, or by other techniques. C. Regulatory Position Emergency core cooling and containment heat removal systemn should be de- =. ~ signed so that adequate. net positive suction head (NPSH) in provided to system pumps assuming maximum expected temperatures of pumped fluids and no increase in containment pressure from that present prior to postulated ' loss of coolant accidents. ~ l MED

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I No. 4 - Discuss the piping runs from the sump in terms of potential air bindage as the flood level rises after a LOCA. The piping runs from the sump to the low-pressure inj ection and Reactor-building spray pumps are designed with a continually negative slope. The reactor building sump is provided with a three foot weir with a top elevation of 87'-6". This weir divides the sump into two sections such that when the reactor' building sump pumps draw the sump level down, during normal operation, the recirculating outlet lines remain inundated precluding any possibility of trapping and/or entraining air in the system as the flood level rises after a LOCA.

== Conclusion:== Because of the large (14") pipe diameters involved in these suction systems and the continuous slope downward from the sump to the pump suctions, there is no location in the system for air binding to occur.

e No. 5 - With regard to the proposal for a 300 gpm flow test from the sump, it does not appear that such low flow rate would be adequate to verify pump suction pressure drops. What is the minimum' flow rate at which FPC would clearly be able to confirm previous head loss calculations (see question #3). Please discuss the rationale'for this conclusion. The intent in testing the sump and suction line friction is to utilize a 1000 gpm flow rate actual and provide direct respective collation to a theoretically calculated value. At this higher flow rate an appropriate differential pressure measuring device (manometer) will be used to measure a ads P of approximately 0.3 psi. In brief summary, the following philosophy will be utilized to correlate the measured 21 P at 1000 gpm to full flow conditions: (1) Create a point and curve utilizing the theoretically calculated value for 1000 gpm cold sump conditions. (2) Create an (actual) curve utilizing the actual zo6 P m'easure-ment at a 1000 gpm flow rate cold sump conditions and project this to full flow conditions. (3) Create a curve utilizing the theoretically-calculated value for a 1000 gpm at hot sump conditions and project ' it to full flow rates.

== Conclusion:== The ar sptance criterion would be that the actual test data and i.s projected curve does meet design requirements. l l

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No. 6 - Discuss test by the manufacturer to confirm the/ho'[uired.

NPSU for each pum building spray). p at Crystal River #3 (decay peat and'JTill j g I9.? n G a;. Attached are the test curves developed by Worthington Co'Ypor ~A j$> s Spray Pumps at Crystal River #3. The performance verification of the Building Spray and Decay Heat Pumps in the recirculation mode with the BWST and at rated flow has been performed in test procedures TP-204-3 and TP-203-3, respectively. Attached are the resultant head capacity curves for the Decay Heat and Building Spray pumps which have been compared to the head capacity curves furnished by the pump manufacturer and found acceptable.

== Conclusion:== Adequate demonstration already exists to confirm pump performance and system performance from data obtained from described tests which have been performed. Points tested confirm ability to measure system conditions at the proposed test flow using sump suction.

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~m No. 7 - With regard to-Table 5 of the December Gilbert study, " potential head elevation available", shouldn't each case read, "from post LOCA water elevation to centerline ~ of ......"? (noting that the top of the sump is at elevation 95 ft.) Under the heading " Potential Head Eley. Available" in Table V of the December 13, 1971, preliminary GAI report, each. case should read "From post LOCA water elevation to centerline of......." This change can be justified by the elevation value shown (99.82') which is the post LOCA water level in the Reactor Building'.

== Conclusion:== Response is "yes". / 4 T-

' eo. ; No. 8 - Since' higher flow rates are desired to verify pump suction pressure drop (see question #5), discuss the feasibility of expandin allow design flow rate'g the capacity of the sump to testing. As an alternative, discuss the possibility of installing temporary piping to permit a pressure drop test at higher flow rates than those proposed (that is, design flow rate of the flow proposed in response to question #5). 4 As discussed in Response #5, the verification of the pump suction pressure drop can be performed at a flow rate of 1000 gpm. This required flow rate can'be maintained from the Borated Water Storage Tank which precludes the necessity for expanding the. capacity of the sump. In addition, expansion of the sump capacity this late in the construction phase would not be practical as much of the: equip-ment and instrumentation surrounding the sump has been installed. Also, the use of temporary piping to allow higher flow rates was considered and determined not viable at this time as the weir arrangement in the sump has been installed.

== Conclusion:== j Adequate data already exists to allow conclusion that 1000 gpm is an acceptable flow for purpose of this demonstration. The proposed demonstration can be accomplished without dis-rupting present construction and/or testing activities in the vicinity of the sump and/or pumps. The ' proposed demonstration can be made without concern for control of the test criteria thus assuring reliable results from a safely cor: ducted test. l 4 I t i l

o No. 9 - It is noted that less than 3 feet margin exists between NPSH required and available (FSAR Table 6 - lla). Discuss the potential that such a margin wo'uld significantly diminish after a LOCA due to ECCS pump flows in excess of design. The ECCS pumps are provided with discharge flow (throttle) control once E.S. signals are bypassed. At the time the operator remote-manually transfers the ECCS pumps cooling source from the Borated Water Storage Tank to the Reactor Building Sump, E.S. Bypass has been accomplished thereby allowing the operator to have throttle control of the ECCS pumps. During this recirculation mode, limit switches of the throttle valves are pre-set to preclude the possibility of flows in excess of design and any resultant effect on NPSH margin. Prior to E.S. Bypass, the limit switches of the throttle valves are pro-set to preclude pump run-out wb.ile ~taking suction from the Borated Water Storage Tank (BWST). No operator action is required during this initial safety injection phase. In addition, in the recirculation mode the flows of the Decay Heat and Building Spray pumps are monitored and alarmed on'either high or low flow conditions. The operator upon alarm can throttle the valves from the control room and thereby correct the high or low flow condition. Operator action in this mode of operation is covered under Annunciator Alarm Procedure AP-102.

== Conclusion:== Both automatic and remote manual control exists in the design to prevent occurrence. Visual indication and alarm are provided in addition. Three ft. of margir. is in Decay Heat Case approximately 14% margin. Other calculation conservativen will in actuality provide additional margin. I l l I -}}

ML19308D692

Text

Dear lir. O'Erlan:

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