ML19308D689

Util Responses to NRC Request for Addl Info Re ECCS Sump Testing

ML19308D689
Person / Time
Site:	Crystal River
Issue date:	10/30/1975
From:	FLORIDA POWER CORP.
To:
References
381, NUDOCS 8003120818
Download: ML19308D689 (30)

Text

{{#Wiki_filter:______ _ ________ 10/30/75 kid 0 tlo UA l9l7$ cMM,3N Additional Information Required on ECCS Sump Testing g-30 For Crystal River Unit #3. INTRODUCTION: We are genera 11y concerned with adequate NPSH, vortex formation and where bindage of the suction. lines. In the case of. Crystal River #3, we note that there is relatively little NPSH margin, therefore, we will require some means of testing to confirm the suction line pressure drop calculations. Most, but not all, of the following questions deal with this area of adequate NPSH.

1. 1 - What is the maximum gravity drain capability to the sump?
2. 2 - With regard to the vortex analysis, are any test data available to support these calculations?
3. 3 - The FSAR provides the following information (Table 6 - 11a).

NPSH Required NPSH Available DH Pump at 3000 gpm 20.0 ft. 22.32 ft. i BS Pump at 1500 gpm 24.8 ft. 22.5 ft. Why do these numbers differ from Gilbert's study dated December 13, 1971? Provide all calculations upon which the FSAR values are based (flood level, velocity heads, elevation heads, head losses, etc.).

1. 4 - Discuss the piping runs from the sump in terms of potential air bindage as the flood level rises after a LOCA.

)g% lh $w"uTBa 6 dN gggg gn 8 o os iso P/fr

8 10/30/75 Additional Information Page 2.

1. 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.

1. 6 - Discuss test by the manufacturer to confirm the required NPSH for each pump at Crystal River #3 (decay heat and building spray).
2. 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.)

1. 8 - Since higher flow rates are desired to verify pump suction pressure drop (see question 5),

discuss the feasibility of expanding the capacity of the sump to allow design flow rate 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.)

1. 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 would significantly diminish after a LOCA due to ECCS pump flows in excess of design. cc: Mr. E. C. Simpson

No. 1 - What is -the maximum gravity drain capability to the sump? i A calculation performed with conservatively assumed friction and head losses resulted in a maximum gravity drain capability to the sump from the Borated Water Storage Tank is approximately 6,000 gpm. System would be without any control to' prevent over-flow of the sump into the containment area. System flow under test. conditions is capable of six times the test flow but without control. Conc 1'icion: Supply for test conditions is more than adequate. i

No. 2 - With rogerd to th2 vortex analysis, are any. test data ,,.available to support these calculations? Portland General Electric Company has performed a sump test for its. Troj an Nuclear Plant and experienced no vortex formation up.to a tested flow of 7500 gpm. The results of this test were filed with ihe Commission in September, 1975. The Trojan Nuclear Plant sump and the sump at Crystal diver #3, although not identical, are similar in that both have 'a weir type arrangement dividing the sump into two sections and both have lorizontal discharge lines of similar size. Attached is a copy of a reference provided by Portland which also ties vortex formation to submergence levels of the inlet piping. The submergence'1evel of the inlet lines at the Trojan Unit is approximately 6.33'. As can be seen.from the GAI vortex analysis, the submergence level of the inlet piping for Crystal River #3 is approximately 6.0 feet (distance from top of sump to pipe inlet). There is also 4.85 feet of water above the top of the R.B. sump prior to the recirculation mode which adds additional submergence margin. The sump test performed at the Trojan Nuclear Plant is not directly applicable, however submergence available under the Crystal River #3 conditions are in excess of those available at Trojan from which we conclude that just based on submergence that vortexing will not occur at Crystal River #3 under LOCA conditions. i

".j ~ ^ 4.. t. .t\\1-3 \\. \\,. s -\\g\\ t 9'o l c .L i . "I . %..- i.l-I 'l '...- v e r i .: :n I }t iQ;;};*.j H ., 4;,:, m, e { y-La (!, i t .: n.. l... ::.....~.. l y,s ?:. ::=a .,, :l.,J.

'("'

i

f. -

.J } si,ld. ne'. turn whi.h.hrp. pu,nse an.) lla.hr .nw liqunl

1. c.. un..-si..n as pi:..q. -u. sp.n I.

.r-sinui.1 t.. nt!.. alo ure } wurt.'in s.l. ins i, hi;:h u l snar;:in of asailable NP. II i-I lim Y litaine inqq.nr. rina pattrin an.1 i e.can.nwn l 8 wl.rir I e }.: i l'.t n..' .:* :.:.it? -~~~-~l s q;e l s.

1. i t

sirh ,._.; q [- - [. f " ' " " l 3: [ c6 ""i i

• c i7:

kI Id. . S. tO.7

f..$

u v..1..k..:.... f3.L:., s, j. ,),.-~.___.___.. .; y v- ? .. tr..:M." :Vo*kI' " ~ .f '} Pung. N... I will pr..hably h.nc tan.,1 >urti..n with casisation g.\\voiil an3 ri>c in en.iion pipe elciati..n whi h can funn an vs. n.1 iil. ration. Pump N. :! runs em. thl. it.1her un re capacity ten iic air in.rkre and make priming and..perati..n imr-.uibic 3 t The suction side of centrifugal pumps When the new pump doesn't perform up to expectations, look for trouble on the suction side. Piping may be inadequate, or its configuration s s may cause turbulence or' air ' pockets. Premature failure then becomes likely h By HAROLD WOODilOUSE, Slovens Institute of Technology .i A nung up rator who i h.ning sinnilc nith a crnini. inn-t in -ri up in peninit ilo innnp in operaic n intrinir.l. i Inpl nnnp luring is, hr.-i fen nn nih-of..perali..n In.nbl Surtion line eie.ing. The puenp nanufa'rturer in ey i ~ j 1 .nh i-r.l in rim. I, -n. iinn. i.le. ne,.liti..n Ihnenn~hly, often -re rir.nl3 n hat i-winn:: 4.n 1 esplain ii I..~icalb.. 'I hi woul.1 ImIp him une.arr the snaj..iity of.!il: cn!!ies Ilm operav.. nr i n tumer. Iln ilm..Inti..e. m.c.1.r n in ini j w hi. b ran ar nl in v.nly pun.p failne.. -neh a-Inni.cn if air n.. liin ation are espen-ise. On. -urh iinati. n i-j haft. noin impellens or u.nu l.c.nin;. w hear a las::e Inunp irpl.n a inailes one. I:nt ilm in sinn l 'nf.n tunatch. an aniin.le r..mn,..n ann ng en:rine ci n..t pipine i Irlt un. hant ril anel now l.c...tne-ina.lc.in..ir. wrli u ne.1 in in.h.mlir i that. a. l..ng a n.e piping \\n..simi te nuble -p.. i. u lmic a n.nsi.n t.. en'..arnes-1 ..mn. l..I 1.. Iln pump - o si..n. it.l...nbl pump.nni, if by n-in -mailer pipe than i-ingnient a nl

p.. it...l. In s nh as. !r - shan espre enl. is n....: I. ilu. pmnp'- f.n:lt.

In.sh of thr e e a-e. in n a ni s cio. in s a i.r-

1. i. s i..n

.). su el:3. the u.. li..n -i.le um 1 1.r well plann..l. Pipr iec h.- e. Inn es - -n. ii..n Ima 1 an 1 pn-ihl>. aa- . :r. ! a..n. nor t !.r 1.u :e riene.h in e ae s. Ihr u. Inn.e. an.1 t!m p.unp li. -nit i-impriler ein-inn. uheation anal hain in..J i. it.cIl inn t f.e pn irm.i s.. mar. h al.e.a alla!.le n.1 p...iiise l' :enp pecileratiini. I nh - gannp .nr . I. f r l I.. -n. tinn 1.c.nl i.N P.-il. In.nhiisi..n. pipe...nli: niation esprai. n. rel engiever. inpo-t-f...po.:ati.n ran b mi--

-OWis

• 6 AY 1/M il

/ /- i X'Q :D RI A ~K' u D., g l d, 3 '- ) .N I. ' i, ->g -]. o g ,s...-.~ - }...,. -_s m -- ~ y [ .g -i 3.--' T-*' m +- ~ f,- { m,. -Y--;=.\\ ~~ p..~ g' Lf; 3 Y fl ~ ,e} y i l dit._. -.ni.... h[r Q I I 1 y . a: m \\

j- -

N foN g:g 4 4 __r \\ N' m e 7 1 %l i 3 x '_s -l( Y vd 'j \\'rrlic I cl!.w pennhr reas..nal.ly ripnl li-tril.ution la inipri. g II..rirontal ellsow distorts and niay rescree fl..w. With d.midc. [ ler. Flow is still Iitter wit! strai I.t tre ti..n 1.cf..rc punip su. tion pump, in!rt en sinall radius *i.lc niay be partit starsetl I t Stvr.:s Sv: tion t Satun d<r A':41 -^ =L s l n,,y y' f I g .__._h i

d

.j ^ }.(tand. int. strai;: lit rednecr allaws an air g.tn air padet c:.n fann wl. crc a syrn. g Eccentric tapered rrducer prnents air k i p.n L. In pisw. anil prienin.: i.lirlicula mrtsical ell.ow appears in curti..n line pr.cLets (leM when pipe >ize is redard i Ir.nlin;r en that NI'.4ll an.1 Statir In a.I lu rnino r..nfu-c.l. e If the w rnarr pennp la e l:<.-en. n that Ihr availal>l..\\l"411 fiorn a tee in a hea.Ier fl?i;. I1. anal fl<.w hernrne-In 1.n. lent. The tinnuph run is nIten lar;cr than the outlet run. i ina.le.guate. li gni.! will ll.. h into v apor. I'.n tiniately, in...: 3.ntng il li.ini.l. air r...l. wl... h in.pnacs th-inai; in ncrtion an ' the li.iniil. whie h inay 1.e n...ving r.:pidis. i-a: ain t fla-hin: an.l c.nisatinn. inial i. to fill the.n allal.!. area in!!y, 't hi4 rau c-a h ral vrincity increa c anel inwerr.1 pre -ure. If the gonnn >er. I'n.l '. en e.f pip-ier :nni inunp peillication c.n> be tion is clo e to ihr ler outlet, th.e Inistnte of tud :n. et clieninate.1 at the le i n sta: e iniply la the pn.per lu. ice li.pii.I an.1 !!#he 1 vagn.r rnay enter the linpel'er eye.1 irm- <.! r.piipinrnt. I en si-r pn.l.lene. anar.l. sel..p..n 11.. ini. IInt c:eniliti..n will 1.r accentuatcel by a further incre ne in s.h. n ti. pip. lin. i. in.e..lf..l. l'ipe rnnli :nrniinn. It i Isne with li.pii1 anil ; a<

s. l.,rits an.1 Ila hin ihn.n-b th.. irnp<llrr.

ll tl.r Innop i. ohiri-e. liv far away fr..in the t c..ani ilnwin:r slunu~h puenps arni run.psr ni-that in.c.an prn. th lie.. i oai-ht enac L. Ih.2 finw will Iraial. ten neI al s i.fr en..u.h pipe r..p.n ilv. l.ut if th. li.pei.I <ir : i-i n..t lst..penly It el up tra II. II.c pip.- w ill I.c..nly pat tially n-. il. till ihr full pipe an.. C..n.r piently, its vclority wi'l i n p. .5 In pic l.orson.fr_ ia

• lew el-....n..n snrai..n i-t, hen e navesfin;; f orue? kine ti.: encryy hark inta pic-nec arni.

a in u.n, in-e.nu c.....II:.a. in. tier v.t.or n~L. t *3'Id .n-r-

b ~ k1..__ () 3 u-

d. '~-....(.

y g, ti a p_y.s. f... 1, .[;., .\\ 4 =l jW La 9

b. -

i. I- ?..... --] ) ......... ~. .e n ..j.

t g-

~.79

• • • • • • • .*j,,.

.,.,,,.,.l .

• t:

~i 3* g, (. c

• • .E

\\,,,.

/.

.. s.. W. . t - ... =......,,r c g. L.. e ( } g l'1..w h..m.ti h.o::r pipe can a. air.l.mn int... ump. If intale J is rl..u, ali p t. in em aion linc. nut Lue r*. ump priformain c j j Sul.nwigr.I.li-.lurge risw.nc 1.laml si;;h mtwrc air mi!! It em Lv.! int.. pipe. : olutirn is to is.ld. insale an.1 ile l. rre j ,l' l ---{- l - j ...n. [. ... n... ..d I i sc g. t h _.... _.. l n l - k1 . j 6 t-i i l' l i {;12 y -.. ~ ---- Et* .n. I I l j L g I ) .l .i .a.

• ..g af.L(J i, c'.Q3 1 j Y4 g

t.; !- - ' - ~. l a b. f w. Le ; O c 4 o 8 ,O 2 .4 b If tm.. turli. Im.s.src Ier. lisas fl rre pipe.li.mitters ap.ut. S=rwnu i sorn ergerice, it lbw 3uths nu) Int. rarl ta ll.c.h niment..I oric or I.oth pisinp-T.. au.i.I surf.wr air i ntrain.wnt, poni.fr greater Inh :.ulv im t;rm e tS. Fi;r.10, or $ or Sn, Fi;r. Ili f..r hi:hrr trha ity 4 6 a rmnbling noi-e an.1 apparent ditinn mi.:ht continue 6.r year, pump vil. ration. This con. as in Fig.1. upper > Letch. but thi.* n.akc4 I wittmrt any pump ilamage ruh. 'I he primin:.lilli. or wear, but the noi.-c ami situation are unapin aling. proper arsangement i< shown in the h.w cr A Y littin; i t'i;. 2 will ilclinitely improse flow hom >ketrh. whcie all p.ut* of she line ringe upwant towant the heailcr. Imt it i-more eslien isc..d tim pmnp without any local mrket.fonning humps, I be u ol in sictermine if thi+ listing is n m.juil;unent must onw 8 mmetiical straizht nr elbow neiluier.- monnic.l hori. 3 m y. Furalwr illo-iration of thi-pains can be -. en in Fiu. 3 zontally al-n perinit alw aurinniilation of air p wket Fig. wlwie suo pump-.fraw (nna the same hemler in parallel.7 aml us. I'rnper meangement has an rcrenaric taperril that there la enmrthing wrong with rnlurer at the >ursion siele with all eccentricity on the The ow nwr enmplain4 pump.L. I-it s ituate<. iti uni y. is elelisers. Imitom i Fig. W. *lims the onesion line will remain full. Ic-than pump.%. 2. > haft-tercak rr mlarly. impeller sam

• wear Suction.l'ne plarenwnt. rather than pipe i t.nt..nr.

aus. etc. Ilut the fault actually lir4 in the >nclinn piptwr. i an lea.1 In air porket* ilne in entrainment..\\ir elraw n int.. .\\nother e sample 14 a hmiennial ellum that holi-ti. the line can leml in Inu of >ursion or imteller en.-inu [ rertly in a pump h..u in-t Fig. 6s. Fl..w n.ight 1.e a4 hum rasitation. In rie.10 ami 11. ihe inv.falline ti-. in. with irincity. \\\\ ascr . harpe ilraps air bubble-into ilm resen nir. 'l he eli.E elicatnl. ami imbulenni im wa-r3 harge temis to pile up on the long.ea.lin. wi.lc. leasing *tnynancy lism-air inn rh.-c to the snrtion inlet anil Imblife-am >m knl in with ahe water. The sointion i< obs inn-.--mus er esen n scr e flaw at tir 3hent raelin< si.le. the pi w* farther ^ap.nt. eninnerpe the lischar e pipe. nr .\\lth..nph thi. e nn.fision may neser f.e notice.1 if the l intespo-e some baille airanermens between them. pump has a rmnfustahl. cap.wity maipin. it may pwsent stelisering rair.1 pump rap;n its. \\luch uni-e. Ihnuph. IhrSursion openin;- shonhl be >ntronn.Inl h3 -ullivient eh.n!.le. mII..n iml+1!rr mumalls I,alann el h hantically water -n that ihr appinaching flow line-remain unir.um. may.lcsclop a steable thin-t an.1 Innu..nl a ilu n-i bear. bail.y nh tinction-m other vnnent-ran can-e eli-t.oini 3 iny. 'Ihe ing-iter : iar fail it..no. nit ti I!nw whh fuehntenre. Iliph hwal schwitic< lea.1 to lla h. .m co.. inn amt imn!.fr w brocser it.. pre arc-in;. po --me in anel casitation. for -month slow. -m rinn the pump m.n h n. a noi > .a ~ If it i m e r-as s to pla.e.m.ll.w nest in rh.. pun p. it line- -In.nl.1 b at lea t 1.~i pipe ili. uncles-In.m ihr -i le. I:.ti.nucser hum the I...itum of the h.ml.I l nn.unt..I sen ticalh e fig..'i.. I hric w ill -liil 1.- m np eI h:. In..nnt .nme li tmei.m..f th.w as. h..w n I.y the it..w lii.e. a liain-tri-aw a3 lin m c.n h othe: i l is.12 hot II..w to boah i.f.i..f the in.prlte, w ;l1 1.e,.unc nraily if th.. infre i-not deep ennngle. Iln month flow line-c.p al. If fra il.le. pla.c 4 ti i he Irweth. m.: h.otri m.n I,.. lu..l.cn at :Lc m la. c...n. tion (n.m is.take-in I c. is.n etian tee.. l su.. n the. ll..w th in iI mm au c w hisip....l -.. that ali th.w-ib.w n the funnel ..mlpmnp,Ihm u ill ti.. n te al-h.cn nar I. f..ve eni. sim. 31..n. ei..n s.a.:.. _...nl si: ht inin ihr inh I. Lies.nnnnil the pipe in 1 :: 12 Air pocket-..m.!rse l..p in a..orte-h lai.I

n. si..n l.c...n.-- sml.nlent an.1.h au s ne t.n c air inin ilw..n tion !

line fr.olie in.m. lak. m I;nr. ' lie teml. m s n -es s nii s.. a pmnp; lhe in. ira e-w ith pipe sch.rin a h m n k e notra. e..r mas 1.c trmptnl in f..ll..= the pinne.1 puaile in i i. l.:. w hi. h lo.w-the mininonn i.rnnunrn.Ini oh nwegem. to asniel m fare.li inesinn. 'R ) j - rowta. mv sua d 0 n

No. 3 - The FSAR provides the following information (, Table 6 ' 11a). NPSH Required NPSH Available DH Pump at 3000 gpm 20.0 ft. 22.32 ft. BS Pump at 1500 gpm 24.8 ft. 22.5 ft. Why do these numbers differ from Gilbert's study dated December 13,~1971 Provide all calculations upon which the FSAR values are based (flood level, volocity heads, elevation heads, head losses, etc.).

RESPONSE

NPSH Required NPSH Available Table 6-11a

• DH Pump at 3000 gpm 20.0 ft.

22.32 ft. BS Pump at 1500 gpm 24.8 ft. 22.5 ft. December 13th GAI Report DH Pump at 3000 gpm 21.66 ft. 24.62 ft. BS Pump (3A) 22.47 ft. 22.72 ft. at 1500 GPM (3B) 23.34 ft. 22.72 ft. June 6 GAI Report B.S. Pump at (3A) 21.10 ft. 22.52 ft. 1500 gpm (3B) 22.34 ft. 22.52 ft. DH Pump N/A N/A 1 l.

• FSAR Tabic 6-11a is from B6W calculations.

l t

No. 3 (centinuad) i Major differences between GAI calculations and BGW calculations (Table-6-11a)-are as follows: ~ BUILDING SPRAY PUMP NPSH CALCULATIONS GAI Dec. 13 Report B6W Calculations NPSHR - 13.0' NPSHR - 15.0' Pump Centerline Pump Centerline Elevation - 78.5' (Revised by Elevation - 77.33' GAI in. June 6 Report to 77.33') R.B. Water Elevation R.B. Water Elevation prior to recirculation - 99.82' prior to recirculation - 99.82' (Revised by GAI in June 6 Report to 99.85') Static Elevation Head - 21.32' Static Elevation Head - 22.5' (Revised in June 6 Report to 22.52') Friction Head Loss - Pump.3A-9.47' Friction Head Loss - 9.8' (used Pump 3B-10.34' for both' pumps) Velocity Head - 1.4' Velocity Head - 0.0' DECAY HEAT PUMPS NP.SH CALCULATI.O.NS GAI Dec. 13 Report BSW Calculations NPSHR - 13.5" NPSHR - 12.5' Friction Head Loss - 8.16' Friction Head Loss - 7.77' Velocity. Head - 2.3, Velocity Head - 0.0' y

~ N3. 3 (centinuad) A c'opy of the B4W calculations with assumptions used to determine =the-NPSH values in Table 6-11a is attached. Copies of the December 13, 1971, and June 6, 1972, GAI Reports are also attached.

== Conclusion:== Refer to FSAR. Table 16-11(A) Table 6-11A POST ACCIDENT NPSH REQUIREMENTS LPI Flow Rate, GPM 3000 Total Elevation Head Available, ft. of H20 22.32 Total Head Required to Pumps, ft. of H20 20.0 (20.27)* Difference + 2.32 (2.05)*

• As constructed - correct values to 2 places.

Pump centerline assumption in error by 0.08 ft. which carried ?.hrough referred to GAI report. 3 ~

T ffggkl3C

• f * **

t E BABC0CK & WILCOX COMPANY ER GENERATION GROUP To l T WALT LICENSING From hhi&36/ (2795) 98.1 DE LEINHART/JR SHETLER -7 UXILIARY SYSTEMS ,,3 3,3,3 st. File No. or Ref. FLORIEA POWER $2bj. Date FORMAL DOL QUESTIONS JUNE 11, 1973 l m. i.., i................. 4.... 6i ..e,. 'In response to your request, Auxiliary Systems is supplying responses to the following DOL questions concerning net positive suction head for the reactor building spray pumps. 6.12 Provide a breakdown of the static elevation head, the friction head los's in the suction piping, the vapor pressure of the fluid and the reactor building pressure, in feet of water, utilized in the post-accident NPSH calculations reported in Table 6-llA for the reactor building spray system pumps. In addition, provide the sump temperature that was utilized in the vapor pressure determination and discuss the relationship of this sump temperature to the sump temperature provided in Figure 14-61, the time building spray is initiated and the time suction is switched from the borated water storage tank to the reactor building sump.

Response

The headings shown in Table 6-11A do not correspond with the numbers listed and should be corrected for clarification to read as follows: Table 6-11A Post-Accident NFLd Requirements LPI RB Spray Flow Race, gpm 3000 1500 To'.. dlevation Head Available, f t of H 0(1) 22.32 22.5 2 TotalHeadRequiredgg) Pumps, ft of H O \\ '20.0 UCEd 24.8 2 Difference +2.32 Cf.od -2.3(3) Notes: (1)Total Elevation Head Available = (Elevation of Borated Water in the RB Sump) - (Elevation of Pump Centerline)

, g, (2) - Tate.1 Hrd R quired to Pump = (Pump Required NPSH) + (Friction Losses n in Pump Suction Lines) V (3) Available reactor building pressure is 11.8 feet at an RB sump temperatura of 211 F (refer to Figures 14-60 and 14-61). Available reactor building pressure = (Reactor building pressure) - (RB Sump vapor pressure). The breakdown of the figures utilized in the post-accident NPSH calculations reported in Table 6-11A for the reactor building spray pumps are as follows: / Static Elevation Head, ft. of H O 22.5 g 2 ((N '. Friction Head Loss, ft of H O 9.8 '

• J 2

RB SumpFluid Vapor Pressure 9 211 F, ft of H O 34.7< g 2 Emactor Building Pressure @ 150*F, [* 1 ft..of H O 46.5 2 V, g The RB sump temperature used for the NPSE calculations is 2110F and was obtained from Figure 14-61 in the FSAR. A further analysis of reactor building spray pump NPSH requirements, at. other post-LOCA conditions, indicates the necessity for throttling. This analysis will be discussed later in this letter. 6 O .13 Describe the reduction in available building pressure utilized in the NPSE calculation of Table 6-11A considering maximum reactor building safety features (e.g. 3 air' coolers and 2 spray systems).

Response

l Table 6-11A considers the reduction in available building pressure due to the 3 reactor building air coolers but does not consider the effect of the 2 spray systems. Further analysis for the purpose of answering this question is not warranted in light of the discussion to follow. In the process of reviewing the reactor building spray pump NPSH requirements, for the purpose of answering the subject NPSH at various points in time between 10guestions, an analysis was rade of 4 see and 106 see after rupture (refer to Fig.14-61). This analysis was based on the assumption that reactor building pressure in excess of RB sump vapor pressuie could be used in calculating NPSH. The results of this analysis are summarized in Table I and 4 shown in Fig. I. Inspection of Table I and Fig. I reveals that between 9x103 see and 5.5x104 sec af ter rupture, there is not sufficient NPSH to prevent cavitation, with 4 the worst case occurring at 2x104 sec. Using the same assumptions that were incorporated in the preparation of Table 1 and Fig I, calculation A was made for the purpose of determining the. degree of throttling necessary to satisfy the RB spray pump NPSHR at 2x104 see after rupture. These calculations indicate that throttling the RB spray pumps to 1200 gpm would satisfy the NPSH requirements. g)vt R NtTm M2l u ./ , g c, Ib) -~

y-Furth r, celculctirn B wcc pr:ptr:d cc tha came eparating conditient po A, but based on the assumption that reactor building pressure in excess of _O

== 14 e6 a1=aee 11==theursu-vere i again, it was found that throttling to 1200 gpm would satisfy the NPSH requirements. In conclusion, we recommend that to insure adequate NPSH available to the RB spray pumps, the pumps be throttled to 1200 gpm at the time of switch-over from the BWST to the RB sump. An NPSH calculation for the DH pumps (calculation C) was also done based A on the assumptions used for calculation B. This calculation indicated there is sufficient NPSH at the design DH pump flow of 3000 gpm for any post-A LOCA condition. DEL /JRS: jgc w/ Attachments CC: CE Barksdale RG Burnley WC Butt LP King JW Merchent EW Swanson 0

t e., 32 BABCOCK & WILCOX 0 6. CERIAt. NUMBER REVJ ' ~ CENERAL CALCULATIONS h fgg gy Oc Puaosc.: _./ o. .le+ee.w w rc e s 9....P p Af_P sS_..._.. p /o w - .....t.c.g ' tc r u.+ A s A +- un i a u1 0 . 'm 4. a : / RS sump nad v&>. 6 peed ec vs. };m e A4'4ev svp+vv<. c urv.<s -f' ear.s Fog.14-61, \\ F Pc. Ps A.R L m e. loss def4 hA. sed ow cnlc /A -/}ou.s b L.&l nz. and L...R kia3 -Ik-FPC y C a l ev 14 4 <.-s - O K6 Speny. Poe,.9 Anh.....Ie. H P.S.W...& / s o o 9 p m... . l> 4><d vaiaus bskas.. \\ . Tnde J~ sammN>.$ai. N eaic.v i0l o ous...:.'... _.- J .J 0 .... Au 4 4.Le. Ms u \\b.s.. 4&.e..1/kb4.cL.iv.. P q.. 1...... $vam Vq.I i4' s ~ fkd-be heaa seen f x1o's ec. s a J.s. r vi o

• see..

4-lau e ss \\ owl sa envah + n ueiaaf M Ps H a 4, p><f.va s+ \\ >ia A nd + L o, F H. e Is ,,a \\ d'Cl<>exce h a.%acu NP.sa n,J.. MP3.Na. \\ s A 4. 2 = t o 4.sae. oce.u s r RB Sp>n Awp hiPiH.) 'l1co. gpn g. xto*.sec-. e l' b 4/h *' I' C T'a s s u r-~p = *z 23. /=. Tga vapa,. i.34 *F P .f 4. 5 2.. g Py ff,Sr l l 1 l a l I f f( 0'. l * 'MI f t e 4 \\O. Pffs P NO. F fl l' NO DWt. Nt ). 4:e *Mf' N*:s ce t t Pl a s..*e s e st t } g,pgoggp No UAIE ) Sill't.I N O. s12L

BARCOCK & WILCOX ~' 00 .D. UERIAt. NUM0l3' REV. GENERAL CALCULATIONS Q _ Hg 2. !bo...ym / 44o=_ y.:......_ ._. _ Y s _ @& M ** 3 e n = ..A Cop _.9pm =

7. ' 8 6 '

_..... A... U/.. t Ng ( 4400 9pn = 7.o.DC- { y $ z. 6 ??,. = -;s 2.7/ ( YS ...Y.f...f.4..oO._.y ...m.-- 2 3_C c 4 //g f /4'00 lffoQffm. 3_' ~ Il as = 22.5,- NPJ/( Pta P ' N Mee ~ V = ff'K2 - 44',32 .9.. /3. ?. ZZ.5 /3.f = O .N. P3l4g. GI4w.3 em c.. L4 54..f bancel ou +>+liaan vp M e+ Eo/ po mp N PSHg ) N esil - N FJ E.I../.3.# U.. rs.'. 5.'.~./.Lat. '. ~ 4-4 n, + cJ.m C n.i% T.~ ~ ~ : te s 7 ~ Pep RPs'ii, { ~/3 = y d T hxso % e Atre> >,,e k a. H 2 1.s-= t 13 a y i N. $. 4 3o o ggu

• 7 0 8 6 ( h ) =

(o. 4 7 z s Hi g. n.ao gy = z.7/ [jkW =

2. 0 4

k. @ in / 4100 y~

9. 7/

i.

• \\ll #f PROP. N O.

F It.h .s ei8 D. W.. s, Ng i. C O *. t. No. sN I N.* e..* *

• I is
• GROtIl' NO.

I' 'O DATE SHEET N.*. ,.. in 1 m,,

~ ..r.....,.,

• ~~ ~'
• %* i^ i '" " " '-
• n......- u.s u n t tiore.

O M Pst4 = qq. s 2 - 44 37 - 8 5/ t-z r. T /f. /z = N P,5 Ag_ g ISoo ya / 4.12-a' a = np.t 4., - Alp.s i4,1 /4. t z - /4. / z. c. -o 'uA n L . kku $4 s'n., $elow As ce.s hsecom A.. 13 ao p,.~ rdS s P &Sk.p,..,4*em N es 6 e /2., y n g z u,- % r ,o }}g k itoa f.42s a.gpa AQ 42oa3p

1. Zos d 17 i

0 }lt [ Iz a $p '2.7/($5Co)'*/73 h/x g n. u //z. =y~ 7/o = de, c 2 ;r,S ' fdPSilp= 4 4.' 52 - 44 3f-7, 9 t ?z, = I4, 7) 11 /A N P.s//g G / t-w pp~.. /3./C D' "\\.S 4 = /4.73 -13. 4L c /,27' NPJ//, @ lLa o yp-i.s s fC c 'J i '/ t-U [.....,.. ...s,. 3

..-.*4,,' i 32 BA8 COCK & WILCOX sa:AL NuMotn n o.i GENERAL CALCULATIONS ( g, C.P, anse s. _.'Th .p...._.._ D s 4.e_.xs_ spen P" ~e _ _....nps_H_s.+._vn.ous cv & t o S, JL%s.a M.my.O E. l *N-f S u *~ y> % ) i b et., _~~ ~_bASA.~ '^ L.w k L.h iscaA en _uGTa~C~u.:U, ~.~_~~.~ ..... l u S 4-.v r.._ e a l 1.P. k.a % T=9C J CsIc ulc-t o>A G) MP3 Ai @ /DPGPM 4,li,y suc9 (Pre " Pv) Auum,.3 A ._... J 's N95 Hg -Ng,

• Nei 1

..._.... 0 A. -

7. 096 + 2 71. =

9

nel, 1)<> sef.' cl:+[n i

dai - ez. c p-NP.sh - 9.i+rr i. ~'. 12, ') '......._.._... _.. _. _. _ _._.. e .__.............._s N. g 3.0.. 61 1..' A = NP.i4 - NBdp.E. h.1.=.LX 1.7_Z,3 ' S h YY Y h, b ...O.D. .M N Psd.,. ~ k + A%. ~ //, Q 4&o gps' '7 04['l2 '- &>,77 //z. ' @ 1400 gpa 2 7)Qfy 2.3G \\ //, g n o ( n d ~ y ~ ~ T ~ $~/3 ~ ~ //a.. z z. 5 e a t i. .I I fh( e ,e ML'i PROP NO. F lt.I' % e tCl s' r DWG N O. C 'M >

• ied osNr NO so.*0 007 GROUP NO.

hgM DA R - A ,A__ SHEET NO. / "',;)./ -llCOS t/,0. /-3 o^'r -/s/73

32 BABCCCK & WILCOX D O C 1.D. EERIAI. NUMUER C EV. CENERAL CALCULATIONS y..._._,-. H Pf//. ;. 2 %I3 +.zz.f ~ A /?,37' = n?SQ @ t4oo u &t =.].5 b.' ps.rnp..NE5.t-lp)(besec{ Va -li a...y."y ew M<t.es. 4 ' /.5,3.7- /4*S~L c - ) ]$..,. N P.s $ [.s no.E//gu,~ lc.~ ~ ~~ ~ ~ ' ~ ~ ~" O.3 M P,5 N @ /S o o g y,. Il. @ ls *.o t'.L3.ea. y n. i llt & 43*ay

5....W % A.$?

.1 o 4 6 1 3 a = y = h,'.[/.'.. [ 3 $, C.* * . 2, o A. 7 o 4 e n.. / 93.? i,k ~ ~.~. ~ b. si ~ C MPk - 9. s~ l + 3 2 5. l.3.99 ....., 1 2-N Ps 4. 81aao yn . /A .o ?.N P1RA. 42.%...' /.3. 99 -/L. I2.? - D. /3 ' ~ ~ ~ s4d ll. d.'KL i~. 5.5.. A. " ' ~.~ ~._~..'._~~ ~. '... ' ~~ ~ ~ ~ 3 3y ht g n... f yz., y s li cp 1z. oo y ' 7.o s[ ' &..-t -- t C.!? // @ / zoo y,.,e

c. 7/

( M Td ' = /.73' t )*. l e iMt it I. PROP. NO FIL.E Nd /e I DWu. No. C 074 8 *. NLs

*4. Na e. H s 'O

{ GTf0tlP NO DA ft Sggg n.,, [ ,O t t H% DATF

....y ~ a:a .,,..,,,..,i:,, orn e e,.* e.r m4: 7T7T: ~- ~i7r T.' i..,.... t r i n o ss 1, n,: w. O Jh. @ lzo o / 4zaoy,.

7. 7 '

//,,- r s.,r ' NPS&. c ~ 7 9 4 ZZ f /At. L = k Ps h. g t&*o yn u 13. 4 4.. g* hg i+. t - i s. s c + \\8 /.it' N Ps4 Q ttoo $pm is s u C.(;c,',J O G e O e e I o l.....,... i t*4:s c'" n

r. m.. 2 a v.....o, ow.. re,- u.m ii..,,,,.. ..i Q,/g, c n. ca. cut.o. riom. o O Rapase Io d.A o m a. b d Par.~p NPS 14 n tyyhameak u4n A. 4.16 s.-p cn 1. l. o n 5 v ( Pgo Py s D,>+a L ib < l.,.r.r c)h cs/es/ b L, sk aa,t L. P.b. mad o&< t=Pc u -m Y k, y, Co lc M,oa r @ A.,,,y ,b y P.>ny NPs#uep,a,Q~,...'g3oae,,m-s e ,o c.~. /,- N PS ll,o 1'Pa.a. = fvp - Pv - & + Wat id#5//g ' -4. F //et 72.s* g 2. x_.g. }let ' % f. 8Z =(22.'EWh_. O q n oy ~

c., z o '

N d.2 coo y = /. T7 t llt{ f9o //suop ' 7. '77 ' 22.32 7,72+ & 2 L ' ;' NPS li - g n, 144 7 /f 55~ ~ - S H PS Ag @ Soaoy r ~ I2 5 A ' MPs Hn - Ni's' & M. Kr 1 ' G4.4b 12,T = (l. 9 7 ') 2. o.r ' ~ ~, w .. Nf6/Ta i.s su-C8'cienf. \\ i F P c_ t-l 'O \\ .<.....-n c. ~: 7 .. u b-t 7.72 ,,, c: <.4 c-10 ', n'hs ' f,s $.30 1] ' ! ' ' *

• f ' ' ^'

~ ~ ~ s ni ' - ff I.4..' e n ? G/,S/73 c -1:$,

. o k..*T ev. V w a t u e)k. .... b... .., I 'F_ _ _ , =. n .t .. L.J. m ....5 .t..s 71 9. g e*.s e ......a is.. ,a -_.I . b'... p, i. r.i.

!e i q,.. _.

t....

..:..s v w -....{..

j i I f l 1: 1 lit L : ~~~ l4 g (;,g ; ,e ~ } g;; ; .p j -g -. j- - -.. I. ? 6 e. e t j g p. 6 e- -r -.d-e-4 -* t - -t e - + - a ..e* ?.. s e ; V.e .1..: : t e ;. -:. (6 ,-j + - - ,O i. 5 g i-tn! !4 j. 'i '. ! ; .j ; : i.I t i m .1 .g, ti t ; ; t. u: I

si

, i I...

..: e ! a:

. i -

IM. f [. . _p. g, I,,. .l. i.[a .i

l.. t,i

-.,. f : j .i. . t.' .t. l. A. 't i i i: \\ Ii'l 'jli!; i 11 l' .i.. ~-lj : 'h 5 $ lj ; I 4.*. -. ....I l l l h I I l I 5I '5 s /,, li j l [. a .w ii

i. :

.: i g I.. h! !I.!,/ I i l. l \\< llI: f o i

1. i

,1 o w J_.. .i. s yre-i;* .F i - .u., ,7 e.. ..si w iti , i. s 'iil til+ N i + !D if!!' i );' i i lil li!i i I A.' t!' tii +.

. /.

.i t jt.! *i ilij. lll[ l !...! i. N. A.t' l iji' 1 1 i CL,. i ! /! i l i i.' lil! !!L :l.h i !!!l it

1. H l':

i N !!!!

t.-

1 m ! !.l . l hl )l '.* * '.. . 'I. ! 3 f.{.i .*'l_. 2 { {.$'..} i. l. ! l ' 0 ' * :..!. I il. C'..... .4. y a l p g ..g;. .e { {.,,

p. ).

6 .t g. 6 a ..{.. J .I ( a t f L i ff. /,... j i

Cr ? -

.-*.1 l; ;t

• t.

. t I i l'ii

• i..f. f,. i !_..i. ' :..

.i;tji.. [Q j j ' 5 ..j[i j ts. 3_ ..i ,i.. 71

t. : b.p; u.4.:etri

q. g.

p L,[..i.6 pi. cr.t., r

: 1-i i-/

ti!! t a c.i. ,t ..i !.;- l'. hI1. O r.s .., : 1 1.. t i.t..+L-5 .:t.4_... .-_f :i:...t..-l...-;j n ,l. i 6 j-. 4 i i p. 4; 6 s e i,. ! l 8 ? r -,. A- -+.6-3 .,.4 i

,g g

f-l 3 -l t } [{f* l-l -[

• g l

I lI1 6 4-li } -l i l l

I.

I1 i i i 1 [f rf ? .f ..;h i-i'e8 o 1 I

• 8 l

I' g t ji -j j i !4 ; y. S - l@ l i i t l! f".;$ ,[ ', I ,{ -l I i l iTen 'f:* i, L i

5

1 Til ri( ifii l'- i r,!! t i '.h !id \\ tit i- ,1 ! i!:: !li iv t f'... \\

8 i

lll, 4 i., r.

! :. s i . /.- m m to w .\\ i.li'...; 1.)

!;i l.i.i.

iiii M ..j / I Lj I'/I.ii!li l!i ill! Illi .i i-lii n.E i '!,! T ;l !! ill i i .lii

j1-s i

i i

1

lit i

..!j .6 J. _.. i t

i; t r' -

,T. i I: i

u. -

t-__: ii! i i.! .ii x.i i A .i }

s tz

,, g/o ! i. i;im = _ _. i /I. !!.: If'!! ! !!9;![iii[I!! ,-l !ili r t.. t-i N!

i i,

gil-. : v l'. it;t I. i..- l-t- ,Is I ;- p t, t i

M 1:1 t'

.' ;r.

: c, 3

,;'l . l

,I

.p,cl l tYi. t.*

1. 5. [. t'.:

i.f_L !.l'l1 l'g.

1...

1 .i i.g { '.L. p l i l : 11 G H'- m / m' l i + j !l ~ ul.9.@q;l ! : 'i LL' y = / a : ax d j !i i j l J M. it/ !i ' i

i

l; l.

a ...,. r...

c... I so u

w w 2 C 4 4 4 r', ;4 @4 n o w i ~ y. s

5,W

s 4 117d

't e .n gg g ;. j,i .i... E,e.t h,i . :..,:i ii r.. 4 c ..q f,; a e W.4 b 'e s a t a b. b 4 bq g 1 o. x i,_.,n o u r w w +. o w , o w a -g ~ ~ g aa e a aa a n a as =a =a >a a n.. .:+ w m n n :~.~ ~,, - n en r; z i = f t ,2*% w w' ! 2 w w. w ' s,,a u wg w G 'w w m n a n ! a g c.~ n i. ', 4,, "4 ' ia g c; 4 g o 4 4 ."n g 4 -j ', a. . o. n.n n. I

o. -

i .i fwfg%' @ = ~ ~ ~. ~. o. ~ ~ n ~ ,a. ~.- r G-M C'. e' ,N e. 0= 6 4 C4 6 6 g4 g j %,,og-s so. 4". 1 ?. 4 .O m w e n m m n-e o w rs m r., 4 4 .e .r m n n v w Q C e, e-g o, 4 .gHg .p q t c an.n :n ,n n .n . n. i ~ ~

1. F e.

U tD H.5 3.k 's c',i[ >2. o 3 4 14 Io h .a & m o o A R A sRe i A. A A,. R. g J l xn =- es. a. o.. i n m w q.. ,i i ll...!.... ,i j i..... .p 1 Ii i l

I

.:-l,' .l: i 6 i i i t i t 1-l l 1, m r-----

<.)3 QUESTION 1.(a). (a) Page 2 of Florida Power Corporation's letter indicates that the vulnerability of dilution Modes I, II, III and IV to single failures is analyzed in Section 10 of Topical Report BAW-10103. Further information is needed from Florida Power to confirm the single failure position stated in BAW-10103. It is the staff's position that Mode I should not be attempted as the primary method to control boron concentration in the core during long-term cooling. The possibility of gas or steam entrainment in the decay heat suction nozzle can result in severe damage to the decay heat removal pump. Long-term heat removal requirements can exist for long durations (days or months) after the accident and continuous operation of one train of the decay heat removal system is required. In the event of an equipment malfunction in this train, no method is available to remove the decay heat if the other train has been previously damaged. Therefore, implementation of Mode I should not be attempted since thh action could result in the decrease of required safety equipment. Since initiation of Mode I is not allowed, it must be established that Modes II, III, and IV in combination are single failure proof. An, alternate to consider is the feasibility of utilizing a single mode, with modifications to make this mode single failure proof and relatively free of pump cavitation problems. We suggest you examine the possibility of maintaining contain-ment sump suction for the LPI pumps, while utilizing gravity drain to the containment sump. To achieve this drainage, small lines may have to be installed from the DHR drop line to the containment sump (with suitable motor-operated isolation valves).

RESPONSE

For Mode 1, it is agreed that the potential exists for some degree of damage to the LPI pump being used. Mode 1 possibly could result in a low level of cavitation which may not necessarily produce a detectable erratic flow indication on the control room flow in-dicator. For this reason, Mode 1 will not be attempted until conditions in the auxiliary building perm?.t a plant operator to be near the LPI pump for pump startup to observe the pump for symptons of cavitation (observe for pump noise and vibration). The modes of operation to be used in long-term cooling to eliminate any potential for unacceptable high boric acid concentrations in the core region are listed below: \\ G2$1 -l'-

Mode 1 - If both LPI trains are operable and a plant operator can be stationed near the LPI pump, try to establish suction ^ from the reactor vessel outlet pipe through the decay heat (DH) drop line with LPI train B. This will force low pressure injection to the reactor vessel to flow through the core. Mode 2 - If Mode 1 was not attempted or was not successful, open the DH drop line to one LPI train to establish gravity draining from the hot leg to the reactor building sump. This will force low pressure injection from the operating LPI train to flow through the core. Mode 3 - If Mode 2 is not successful, open the auxiliary spray line to the pressurizer. This will route dilute injection to the area above the core. The flow path is through the auxiliary saray line into the pressurizer, out of the pressurizer through the surge line into the hot leg and then into the reactor vessel. Modes 2 and 3 in combination are single failure proof. Mode 2 (gravity drain to sump) only utilize the suction side of one of the two LPI trains. Mode 3 (auxiliary spray to pressurizer) can be provided by either of the two LPI trains. The response to question 1. (c) and the single failure analysis below of individual motor operated valves demonstrates that Modes 2 and 3 in combination are single failure proof. Valve Description Failure Evaluation DH drop line valves Would prevent Modes 1 6 2, (DHV-3, 4 or 41) Closed Mode 3 unaffected Emergency sump outlet Would prevent Mode 2, Valve (DHV-42 or 43) Closed Modes 1 6 3 unaffected LPI flow control valve Would prevent Mode 1, (DHV-14 or 25) Closed Modes 2 6 3 unaffected LPI injection valve (DHV-5 or 6) Closed No effect Aux. spray valve Would prevent Mode 3, (DHV-91) Closed Modes 1 4 2 unaffected Pressurizer main spray No effect. For Mode 3, line control valve (RC-V1) Open RC-V3 is closed, which accomplishes intended function Pressurizer main spray No effect. For Mode 3, line block valve (RV-V3) Open RC-V1 is closed, which accomplishes intended function. 3 QUESTION lb: Provide the specific operating procedures required to implement each of these dilution modes. Indicate each operator action necessary, and identify which valves must be manually operated outside the control room. Confirm the capability for the operator to close or open these valves at the time specified in the September 19th letter (to be instituted 15 minutes after a LOCA) and evaluate the radiation levels which could be present at the valve locations.

RESPONSE

The specific plant operating procedures required to implement the dilution modes are presently being developed.. However, once completed, the plant operating procedures will include the procedures outlined below. Procedures. The ECCS systems shall be placed in one of the following th.ree operating modes within 24 hours after the accident. Injection flow to the R.C. systsms should be maintained through two paths while attempting to place the systems in one of the three operating modes. The two inj ection flow paths can be either the two LP injection lines or one LP injection line combined with one HPI string (LPI pump acting as booster pump for HPI pump). Mode 1 - Attempt to Establish Suction from Hot Leg with One LPI String. a. This mode shall be attempted only if both LPI trains are operable and if conditions in the auxiliary building permit an operator to be stationed near LPI Pump 3B. If this mode is successful, it is indicative that the R.C. System is filled with coolant to above the hot leg elevation. LPI Train B is to be used to take suction from the hot leg. b. Open DH drop line EMO valves DHV-3 and DHV-4. Since LPI Pump 3B will be shut off for the DH drop line c. suction attempt, obtain inj ection flow to the reactor vessel through two injection flow paths by either of the following: (1) Place LPI Train A in " piggy back" with HPI Train A. Open LPI to HPI crossover EMO Valve DHV-11. Start HPI pump in Train A and use HPI EMO valves'MUV-23 and MUV-24 to control HPI flow to 500 GPM. Use LPI EMO Valve DHV-14 to control LPI Pump 3A flow rate to 3000 GPM (LPI pump flow is the sum of LPI Line A flow rate and HPI Train A flow rate). (2) One LPI pump operating with the flow split between the two LP injection lines. Open the LPI to LPI crossover manually operated valves DHV-7 and DHV-8. Shut off LPI Pump 3B in Train B and close Train B EMO flow control Valve DHV-25. Use LPI EMO valves and DHV-5 and DHV-6 to split LPI Pump 3A flow between the two LP injection lines, approximately 1500 GPM. per line. Flow Measurement String 1 should indicate 3000 GPM and Flow Measurement String 38 should indicate 1500 GPM. d. Shut off LPI Pump 3B, if not already done, and shut off the building spray pump connected to the same suction line as LPI Pump 3B. e. Close Sump Outlet EMO Valve DHV-43. Close LPI Train B Flow 1 Control Valve DHV-25. i f. Open manually operated Valve DHV-40 in the DH drop line. Then open EMO Valve DHV-41, also in the DH drop line. g. Station an operator near LPI Pump 3B to observe pump for symptoms of cavitation (observe for pump noise and vibration). h. Start LPI Pump 3B in Train B and slowly initiate and increase flow by using EMO Flow Control Valve DHV-25. If cavitation symptoms develop, stop increasing the flow rate and reduce it slightly until the symptoms cease. If the symptoms do not appear, stop the flow increase at 3000 GPM. LPI Pump 3B is now taking suction from the hot leg only. i. If-HPI Train A is in " piggy back" operation, it may be shut down. LPI Train B is now taking suction from the hot leg and pumping to the reactor vessel; LPI Train A is taking suction from the R.B. sump and pumping to the reactor vessel. j. An additional step may be taken, when convenient, to deternine if the break location is high enough in elevation to operate only'one LPI Train with-suction from the hot leg. Slowly decrease the flow rate in LPI Train A and then shut off LPI Pump 3A in String A; continuously observe LPI Pump 3B for symptoms of cavitation. . Coolant from the sump is not being 'i 2-

pumped to the reactor vessel now; i.e., not providing an overflow out the break. If the suction to LPI Pump 3B is not lost, it is indicative that: (1) the R.C. System is filled to above the hot leg elevation, and (2) the LPI Train B Inj ection Line is intact. LPI Train A may now be placed back in operation (taking suction from sump) or operated periodically to make up for volume contraction as LPI Train B reduces the. reactor coolant temperature, k. If Mode 1 was not attempted or was not successful, Mode 2 will be attempted. Mode 1 could fail because the water level in the hot leg was not high enough to prevent steam or gas entrainment in the DH drop line or the valve disc for one of the DH drop line EMO valves did not actually lift. Mode 2 - Open the DH Drop Line to the R.B. Sump a. If Mode 1 i.s not =uccessful or was not attempted, maintain injection flow to reactor vessel through two inj ection flow paths by one of the following: (1) One LPI pump operating with LPI discharge cross connect open (manually operated valves DHV-7 and DHV-3) and the flow split between the two inj ection line s by throttling EMO Valves DHV-5 and DHV-5. (2) One LPI Train in " piggy back" with one HPI train. b. If one LPI pump is not operating, use that pump's suction piping for Mode 2. It is preferable that'LPI Pump 3B be shut down for Mode 2, if there is a choice. c. LPI Pump 3B shutdown. BS pump associated with same suction line as LPI Pump 33, shutdown. DH EMO Flow Control Valve DHV-25 closed. R.B. Sump Outlet EMO Valve DHV-43 closed. DH Drop Line EMO Valves DHV-3, DHV-4 and DHV-41 open. DH Drop Line manually operated Valve DHV-40 open. d. R. 3. sump outlet EMO Valve DHV-43 is the only closed valve in i.5e flow path now. The temperature measurement in the DH drop line will be used to determine whether the drop line is draining by gravity to the sump or not, because is is possible that a valve in the flow path is not actually open. The indicated temperature should change if gravity draining exists. Note the DH drop line indicated temperature and then open sump outlet EMO valve DHV-43. A change or a fluctuation in the indicated temperature is an indication of flow. If Mode 1 was not attempted, then the initial temperature indication will be low (auxiliary building ambient temperature) and gravity draining from the hot leg will significantly increase the indicated temperature. i I L f 1 If Mode 1 was attempted and not. successful, but coolant from the hot leg reached the temperature sensor, then the initial temperature indication could be high when the gravity drain is attempted. If this has occurred and if when gravity draining is attempted, the indicated tempera-ture does not change, then the following should be done: (1) Line-up a flow path from the operating LPI Train (Train A) through the crossconnect into LPI Train B, backwards in Train B, and backwards through the DH Drop Line to establish a different indicated tempera-ture in the drop line._ DH drop line valves DHV-3, DHV-4 and DHV-40 and Sump Outlet Valve DHV-43 are still open. (2) If the crossconnect (Valves DHV-7 and DHV-8) is not open between the two LPI Lines, open it. Close sump outlet EMO valve DHV-43. (3) Close LP Injection EMO Valve DHV-6 if this injection line is not being used for injection to the reactor vessel. Partially open LPI EMO flow control valve DHV-25. This has now established a backflow path through the DH drop line. (4) When the cool discharge fluid from the DH cooler establishes a new indicated temperature for the DH drop line, stop the backflow by closing EMO Valve DHV-25. Re-establish the gravity drain flow path from the hot leg to the sump by opening sump outlet EMO Valve DHV-43. A change in indicated DH drop line temperature confirms the existance of gravity draining. Maintain inj ection flow to reactor vessel through two injection e. flow paths per Item a., above. Mode 3 - Open Auxiliary Spray Line to Pressurizer This operating mode will be used if both Modes 1 and 2_are a. unsuccessful. b. Close main pressurizer spray line EMO Valve RC-V1 or RC-V3 or both, t I _

i i If LPI Pump 3A is operating, open auxiliary spray line' c. EMO Valve'DHV-91..The spray flow path is now established. Opening and closing EMO Valve DHV-91 will produce slight changes in the indicated LPI Train Flow Rate (Instrument String 1) to confirm that spray flow exists. o I d. If LPI Pump 3A is not operable, the crossconnect (manual j valves DHV-7 and DHV-8) between the LPI trains must be opened :to provide a flow path from operating LPI Pump 3B to the auxiliary spray line. Backflow through the idle LPI Pump 3A must be prevented by closing the pump l suction manually operated Valve DHV-21 or by closing sump outlet EMO Valve DHV-42 and BS pump su; tion EMO Valve BSV-17. Maintain injection flow to the reactor vessel through e. two injection flow paths per Item a. of Mode 2.

---

End'of Operating Procedures -----------------

The manually operated valves involved in the above procedures are as follows: Mode 1 Mode 2 Mode 3 i DH Drop'Line Valve DHV-40 Yes Yes LPI to LPI Crossconnect DHV-748 Yes(l) Yes (2) (1) If the second inj ection path to the reactor vessel is the LPI line. (2) If LPI Pump 3A is not operable. The operator is allowed 24 hours to implement one of the three operating modes rather than the 15 minutes time period mentioned i in the September 19 letter. Once it has been established which of the dilution modes will be implemented, the two eight (8) inch LPI crossconnect' valves, DHV-768, and tha fourteen (14) inch DH Drop Line Valve, DHV-40, are readily acces ible to the operator in the Auxiliary Building. The radiation level is greater than 100 mr/hr (Reference GAI Report No. 1824 Shielding Design-Report t for Crystal. River. Nuclear Generating Plant' Unit No. 3), however, the operator exposure time is very short term and the radiation levels are therefore not considered a pertinent factor relative j to valve operation. -Eo

I e a L O a. i

i e

%.49l[ 9 I--N N ,;n

• C l

,e k[fkIf l,, N. il; 'a N Ag -[ a i 1 8 Avi s i... --N N 09-C 3-Sh 9 x tu i Q rgg ;; A.: s xil. s .i I--N N = if 3 ir 1 [ 'E I 1l, I N 1-i ,l d g ,d$

d3, I

l ,o ~- N e w -a g i v.

1,

t _. j, ) 6)-"1! l l 18 g, Q,3 8

:3:

5 1 g 3- -s .a n o N ~ \\,,7 F" h 'i.v'!,M., g'! (3, G3 j = i i ~ T J t i {#[, [I!() ;'. d- ' N >i 5 ~ I _t l y a i s!! ~ T,h 9 e

QUESTION 1. (c) Florida Power Corporation indicates that a backup electrical supply is present to run the pumps and actuate the motor-driven valves for all modes in case of electrical failure (see reference to BAW-10103). List all pumps and valves included in the. dilution modes with their associated power supplies. 4 Specify their power requirements and relate to the capacity of the backup power supply. Indicate the location of this power supply in the plant.- Specify those motor-operated valves in the dilution modes which, if rendered inoperable due to a common power supply ' failure, would fail all dilution modes. Discuss your means to connect power to these valves should they be inside containment.

RESPONSE

The following table lists the valves and pumps associated with the various dilution modes, their channel, power supply and power supply location: Power Supply Mode Valve / Pump Channel Power Supply Locations 1 LPI Pump 3A A 4160V ES Bus 3A Cont. Comp.Elev.108' LPI Pump 3B B 4160V ES Bus 3B Cont. Comp.Elev.108' i DRV-3 A ES MCC 3Al Aux. Bldg.Elev. 95' DHV-4 B ES MCC 3B1 Aux. Bldg.Elev.119' DHV-41 A or B(2) ES MCC 3AB Aux. Bldg.Elev.119' DHV-43 B ES MCC 3B1 Aux. Bldg.Elev.119' DHV-25 B ES MCC 3B1-Aux. Bldg. Elev.119 ' DHV-6 B ES MCC 3B1 Aux. Bldg.Elev.119' DHV-5 A ES MCC 3Al Aux. Bldg.Elev. 95' 2 DHV-3 A ES MCC.3Al Aux. Bldg.Elev. 95' DHV-4 B ES MCC 3B1 Aux. Bldg.Elev.119' DNV-5 A ES MCC 3Al Aux. Bldg.Elev. 95' DHV-6 B ES MCC 3B1 Aux. Bldg.Elev.119' DHV-25 B ES MCC 3B1 Aux.31dg.Elev.119' DHV-41 A or B(2) ES MCC 3AB Aux. Bldg.Elev.119' DHV-43 B ES MCC 3B1 Aux. Bldg.Elev.119' LPI Pump 3A A 4160V ES Bus 3A Cont. Comp.Elev.108' 3(A) LPI Pump 3A A 4160V ES Bus 3A Cont. Comp.Elev.108' DHV-42 A ES MCC 3Al Aux. Bldg.Elev. 95' DHV-91 A or B(21 ES MCC 3AB Aux. Bldg.Elev.119' RC-Vl(3) A or B(2) ES MCC 3AB Aux. Bldg.Elev.119' RC-V3(3) or A or B(2) ES MCC 3AB /ux.B12g.Elev.119' A or B(2) ES MCC 3AB Aux. Bldg.Elev.119' RCV-53( )or DHV-14 A ES MCC 3Al Aux. Bldg.Elev. 95' DHV-5 (4) A ES MCC 3Al Aux. Bldg.Eley. 95' no m

s Power Supply Mode Valve / Pump Channel Power Supply Locations 3(B) LPI Pump 3B B 4160V ES Bus 3B Cont. Comp.Elev.108' DAV-43 B ES MCC 3B1 Aux. Bldg.Elev.119' DHV-5 A ES MCC 3A Aux. Bldg.Elev. 95' i DHV-6 B ES MCC 3B1 Aux. Bldg.Elev.119' DRV-91 A or B(2) ES MCC 3AB Aux. Bldg.Elev.119' RC-Vl(3) A or B(2) ES MCC 3AB Aux. Bldg.Elev.119' RC-V2(3) or A or B(2) ES MCC 3AB Aux. Bldg.Elev.119' RCV-53 A or B(2) ES MCC 3AB Aux. Bldg.Elev.119' NOTES: (1) Mode 1 will be attempted only if both LPI trains are operable and only train B would be used for Mode 1. (2) Manual transfer switch provides capability of power supply from power train A or B. (3) Only one valve is required to be closed. (4) Only one valve is required for throttle. If power is not available to a valve that must be operated, the valve can be operated by one of the following methods: (A) Operate valve using local hand wheel on the valve operator. (B) Restore power to the bus serving the valve (In the event one power train is lost, backup power can be supplied from the remaining power train by use of the cross connect features as described in Sections 8.2.2.4 and 8.2.2.5 and depicted on Figures 8-7 and 8-9 of the FSAR.) Based on the above, there are no valves which, if rendered-inoperable due to a power failure, will fail any dilution mode. Supplying backup power to the valves is not effected by location, as.the supply busaes~are located external to the containment. 1 I i __ l

QUESTION 1.(d) Confirm that remote readouts of dilution flow rates for each mode are available in the control room. _ RESPONSE. Remote readouts of dilution flow rates for Mode 1 is available in the control room. For Mode 2, temperature measurement in the DH drop line will be used' to determine whether the drop line is draining by gravity to the sump. For Mode 3, using Train A of the LPI system, opening and closing EMO valve DHV-91 will produce slight changes in the LPI Train Flow Rate Indicator (DH-001-FI 1) which is readout in the control room.. These changes in flow will confirm that flow exists in the auxiliary spray line. For Mode 3, using Train B of the LPI System, local flow rate indication is available at Flow Indicator DH-038-FI 1 in the crossconnect line between the two LPI trains. 1 l O i

QUESTION 1.(e) Discuss the capability to test the dilution modes.

RESPONSE

The operability of the lines, valves, and pumps of the systems used in the dilution modes identified in response 1.a. 4 1.b. is_ verified during pre-operational tests: (1) TP-600 Nuclear Services Closed Cycle Cooling, Nuclear Services Seawater, Decay Heat Removal, Decay Heat Closed Cycle Cooling, Decay Heat Seawater - Operational Test. (2) TP-500 Reactor Coolant Chemistry Test. (3) TP-203 Decay Heat Functional Test. (4) TP-310 Reactor Building Sump Test. The auxiliary spray line and the decay heat drop line portion of the dilution modes are utilized during. normal cooldown of the reactor and the systems utilized in the dilution modes that are common to the ECC system are required by the Technical Specifications for Crystal River #3 to be periodically tested. i

QUESTION If: Explain the type of valves that DHV-39 ar.d DHV-40 are, and identify their mode of actuation.

RESPONSE

Type: Gate valve, bolted bonnet Manufacturer: Chapman Material: A351-CF8M: Stellite seat End Conn.: Butt weld Working Press.: 300 psig t'orking Temp. : 300*F Size: 14 inch Press: 300 lb. Pipe Code: B31.7 N-2 Seismic Class: I Handwheel operated with handwheel extension. ) L

i ' QUESTION Ig: ) t Provide the elevations of the piping and other components in the decay heat drop line from the hot leg nozzles through each of the trains to the reactor building sump. This information will verify that gravity draining for boron dilution for Mode II is possible.

RESPONSE

As indicated on the attached isometric sketches (No. I thru No. 3) the DHR drop line has essentially a continually descending elevation from RCS (Eley. 128'-0") to the Reactor Building Sump (Elev. 86 ' - 3"). Hence, providing adequate gravity draining for boron dilution for Mode II. 6 1 4 1 l

Nu. I .e PEN # 344 DHVK 39 I I I

4. El.8sC" W.P. El.18 7 '-O"

\\ DHV-3(F) W. P. E l. ll 5 '- O"---+- PE g 344 DHV-4I(F) W.P. E 1.106'-O"" 4 L El. 8 6'-6" 1 i El.103 6"" R.B. SPRAY PUMPS DHV-40 SSV-17(F) L

• t El.76 6" E E l 8 8 '-O""

!(81*-6" . E I. t El. 86'-6" R.B.SP PUMPS \\*t El 76'-6"

2:) r Ng,2 ~. PEN-#345 Reactor Bldg.

t. El. 8 s -3" Sump PEN #345 DHV W. P. E l. 85 '- 2" 43(F)

W. P. El. 84'-6 %" 3 W. P. El. 8 5'- 0 % " W. P. EL 84'-6 /s "

f. E L. 8 2* 2 % "

W. P. El. 84*-9 /je" 3 W. P. E 1. 8 4'- 3'/e" R.B. SPRAY PUMPS BSV 16(F)

i. El. 76'-6"

) -}, N3.3_ P EN # 346 Re ctorBld.

4. El. as' 3" 9

/ i PEN #346 DHV 42(F) W. P. E l. 8 4'- g g" W. P. Ef. 85'-5 /,a 3 l ke " ~ J W.P.El.85*.O f," I R.B. SPRAY - PUMPS BSV 17 7 ) hi N( Ei.76 6-T e I

QUESTION 1.(h) Mode III: Justify the 40 gpm predicted to occur through the auxiliary pressurizer spray line. Since water is being pumped from the containment floor, discuss the ability of the spray to pass all material admitted through the sump screens.

RESPONSE

The driving head to produce 40 GPM minimum flow through the auxiliary spray line is essentially the pressure in the LPI line at the take off point for the auxiliary spray line. The LPI pump has a design TDH of 150 PSI at the design' flow rate of 3000 GPM. Of this 150 PSI, 100 PSI is for reactor vessel pressure with the remaining 50 PSI available for A P in the LPI lines. The design TDH is based on the LPI pump taking suction from the BWST (atmcspheric pressure) and pumping 3000 GPM into the reactor vessel with 100 PSIG pressure in the vessel. For Mode 3, the pressure in the containment and l the reactor vessel will have equalized and the " extra head" of the pump will be taken as pressure drop ac'ross LPI control valve DHV-14. This will result in a pressure at the auxiliary spray line take off point which is sufficient to produce more than the required 40 GPM minimum for dilution in Mode 3. The water pumped from.the containment passes through the containment sump screen, which has 0.25 inch openings, before entering the LPI system. The spray head in the pressurizer has a 2 inch orifice diameter. j -e-e- ,--,---,~r-en,,,, e-,--- e e-,-~m,w-e --.m- ,r--v r

' v. QUESTION li: Indicate the feasibility of monitoring boron concentration levels during the long term.

RESPONSE

In systems utilized for boron dilution flow the following sample connections are available: 1. Decay Heat (LPI string A 6 B) utilized for all modes CE-131. 2. Pressurizer water space CE-125. l l l ~..

QUESTION 2. 4 (a) Confirm that a single failure or operator. error that causes any motor-operated valve to inadvertantly actuate could not adversely ' affect the ECCS (i.e., Service Water System Valves, Building Spray System Valves, Boron Dilution Valves, CFT Vent Valves, etc.). (b) Provide a list of each valve considered under item a) and indicate the consequences of the spurious actuation. RESPONSE. ECCS Single Failure Analysis - General. 1. The operator cannot override an ES control signal to an individual component. 2. Power to buses ES-A 4 B are independent such that one train will remain operable even if the other train fails completely. 3. Flow Diagram

References:

FD-302-601: Nuclear Services Closed Cycle Cooling FD-302-611: Nuclear Services and Decay Heat Sea Water FD-302-631: Decay Heat C1 coed Cycle Cooling FD-302-641: Decay Heat Removal PD-302-651: Reactor Coolant FD-302-661: Makeup 4 Purification FD-302-673: Nitrogen and Hydrogen FD-302-702: Core Flooding System ~ 4. This analysis is performed in accordance with Items 2(a) and 2(b) of the NRC request for additional information attached to the December 8,1975 letter.- ,~ ~.., -,-.y 1 e s' .' y e h e o i e an

i ECCS Single Failure Analysis - HPI Actuation - General. 1. One HPI pump is sufficient for HPI cooling per Tech Specs. 3/4.5 Emergency Core Cooling Systems (ECCS). 2. There are two separate ar.d independent flow paths in the HPI flow path to the reactor so that a single failure in one path would not prevent the other path from providing the necessary cooling. Also in the return lines to the makeup tank from the RC system, there are series isolation valves so that a single failure of a valve closure will still prevent a flow path from existing to the makeup tank. 1 j p f,- S* e O -.i g g 6. l e

1. h'.*f >.

e,

. }* 1

];. l .... - a ~. 3, . o e g O 9 8 4 s = - ,7., ru <.4 , 4 4p .,.d,,. .= ~ e 's -? -.._...L.i des S -. n..-e - .,2,. a.' - +

BCCS Singlo Fdlure Analysis - HPI Actuation Sing 19 ~ Failure or Component Operator or Event Controlled By Action Required Error Inumamt

1. HPI Valves to Reactor Open on HPI One valve fails to open or Sufficient flow will be provid Inl:t Lines:

Actuation one ES train fails 4 two by redundant flow path. valves in same loop fail to MN-23, loop A ES-A open. MN-24, Icop A ES-A MIV-25, loop B ES-B MN-26, loop B ES-B

2. HPI Pumps ES-A 4 B A second ptmp in

a. Either of the two ptmps

a. Punp that does run will addition to one fails.

MJP-1A provide sufficient flow. nonnally operating De two selected pumps are MJP-1B punp starts on

b. Operator has incorrer.tly valved such that each feeds MIP-1C HPI Actuation.

selected or valved nor-one independent flow path, mally operating punp and "ES Standby" ptsap such

b. Operating Procedure OP-402 that only one flow path does not allow this.

is being fed. However, sufficient flow

c. Cooling water to one ptznp would be provided through fails rendering that pump a single flow path, inoperable (i.e., Failure of nuclear services closed

c. Remaining pinps are cooled cycle cooling or one train by separate 4 independent of redundant decay heat paths from the one that closed cycle cooling sys-failed such that the req'd tem).

second ptap will remain operable. Eus, one ptmp

d. Ioss of cooling capability will provide the req'd flowc to both the Nuclear Ser-vices and Decay !Ieat

d. H is would render all HPI Closed Cycle Cooling ptmps inoperable. However, Systems as a result of the loss is considered not bearing failure in the credible, because the piping five sea water punps due is seismically qualified 5-8 to loss of flush water pipe, low tenp. (ambient),

from a passive line fail-low pressure (150 psig max.) ure in the discharge side 4 low energy fluid piping. of the flush water booster R us, cooling capability is pumps (DOP-2A 4 DOP-2B). not affected.

ECCS Sing 19 Failure Analysis - IFI Actuation (Cont'd) Sing 19 Failure or Component Operator cr Event Controlled By Action Required Error Comment 3, IDCA Area is small Operator Operator must line Operator fails to start Operator will receive a IFI causing IFI to deplete up proper valves recirculation. 10w flow alana at which time' Borated Water Storage for Reactor Bldg. he will have to start stap Tank before LPI is Stmp recirculation recirculation to provide started. through LPI Ptaps required flow. Failure to to IPI Ptaps. heed this low flow alarm would be a second error. Rus, cooling capability not affected.

4. Borated Water Storage Open on HPI One valve fails to open.

Parallel valve will provide Tank to Ptep Suction Actuation required flow to one ptmp string. Isolction Valves: MW-73 ES-A MW-58 ES-B B. Meksup Tank to Close on IFI Valve fails to close. Bulk of ptmp suction would Makeup Ptap Iso. Actuation normally be from BWST. Any Valve: suction from makeup tank would be limited to the tank inventory MN-64 ES-A 6 B which is nonna11y 2250 gals. Feed to tank is cut off upon IFI Actuation. Rus, one IFI string would not be affected 6 provide the necessary flow. He ~ other string would also provide flow with a makeup tank /BWST mixture for a short period of time not affecting cooling capability.

3. Imtdown Cooler Iso.

Close on Reactor Valve does not close. Upstream valves MW-40, 41 Valva: Building Press. controlled by ES-A will close > 4 psig. to prevent flow. MW-49 ES-B

BOCS Singlo Failure Analysis - EPI Actuation (Cont'd) Singlo Failure or Component Operator or Event Controlled By_ Action Required Error r-nt

7. Intdown coolers Close on Reactor Q1e or both valves do not Down stream isolation valve MN-d MllE-1A 4 MHE-1B Bldg. Press.

close. controlled by ES-B will close Iso. Valves > 4 psig. prevent flow. MN-40 ES-A MN-41 '8. Seal Bleed Off Iso. Close on Reactor Valve does not close. Upstream block valves in individi Valve. Bldg. Press. purnp return lines closed by ES-L\\ > 4 pdig. to prevent flow (MN-258, 259, MN-253 ES-B 260,261.)

9. Individual pump Close nn Reactor One or more valves do not Downstream isolation valve seil bleed off iso.

Bldg. Press, close MN-253 controlled by ES-B valves: > 4 psig. will close to prevent flow. MN-258 ES-A MN-259 MW-260 MN-261

10. Pimp Recirc. Valves Close when Reac.

One valve does not close. Since the valves are in series to M2eup' Tank: Bldg. Press. the other will close to prevent > 4 psig. recire. of borated water to MN-53 ES-A maketp tank. MN-257 ES-B '11.Mik:up Isolation Close when Reac. Valve fails to close. Provides another parallel path Velve to Reactor Bldg. Press. for IPI. Does not affect coolinj loop A > 4 psig. MN-27 ES-A 4 B

ECCS Single Failure Analysis - WI Actuation - General, 1. Che LPI punp and flow path is sufficient for LPI cooling per Tech Specs. 3/4.5 Emergency Core Cooling Systems (ECCS). 2. 'Ihere are two separate and independent flow paths in the LPI flow path to the reactor so that a single failure in one path would not prevent the other path from providing the necessary cooling. i l ) i 1 i l l l

' EOCS Singic F2ilure Analysis - LPI Actuation Single Failure or Component Operator er Event Controlled By Action Required Error Comment (Components or Events Affecting IPI from Both Borated Water Storage Tank (BWST) Suction and Reactor Building Bnergency Simp Recirculation) 1. LPI Bnergency Open on LPI One valve fails to open Parallel valve will provide Injection Valves: Actuation required flow from one ptmp string DW-5 ES-A 11W-6 ES-B 2. W Removal Heat Open on LPI h e valve fails to open Parallel valve will provide Exchanger Discharge Actuation required flow from one ptmp Valves: string HW-14 ES-A 11W-25 ES-B 5. DH Removal Heat N e heat exchanger fails to Parallel flow path will provide Exchangers: provide cooling. necessary cooling and flow DfE-1A DiffE-1B l. Deccy Heat Pumps: Start on LPI One ptmp fails. Parallel pump will provide Actuation required flow. 111P-1A ES-A NP-1B ES-B

i. Any passive element Failure of any one in either

% e unaffected flow path will in m System flow path, provide necessant flow for cooling.

i. Cooling Water to Hi Failure of one DH CCC flow Since the DH CCC is also redtn-Removal Heat Exchangers dant, only one heat exchanger and DI Pumps IU Closed and ptmp is affected. Rus,

Cycle Cooling (CCC) req'd cooling flow will be pro-System vided by the unaffected LPI pathj 4

BCCS Sing 19 Failure Analysis - LPI Actuation (Cont'd) Singlo ~ Failure or Component Operator cr Event Controlled By Action Required Error Comment 7. Di 4 Nuclear Service Total loss due to failure of h is would result in inability Sea Water Cooling all sea water pumps for loss to remove heat from Di System. System (Ultimate of flush water from flush Passive line break is considered, Heat Sink) water booster ptmps due to not credible because pipe is passive line break. S-I pipe, low temperature and im pressure. (See HPI Actuation Analysis, 2d.) R us cooling capability not affected. B. Di Ptnp Discharge' Operator Normally Closed Operator opens one valve. thaffected ptmp string will Isolation Valves to Only provide required flow. Affected Mak:up Pre-filter: string will also provide flow but at a redwed rate. DW-105 DN-106

7. ~ Hi Isolation Operator Normally Closed Operator opens one valve at Unaffected LPI ptap spring will Valves to Makeup Only incorrect time, supply required f1w. Affected (HPI) Ptaps Suction:

pump string will supply reducted flow rate; however, makeup ptmp HW-11 will return the suction flow HW-12 from Hi System to the Reactor Coolant to provide cooling.- LO. RC to Hi Isolation All are to remain One valve is opened either he other two valves remain Valves: closed until RC by ES or operator when RC closed preventing crossover pressure is pressure is above crossover flow since these three valves H W-3 ES-A below the de-setpoint. are in series. Rus ECCS UN-4 ES-B sign pressure function is not affected. NW-41 Operator of 31 System. Ortly .1. Nakttp Filter Iso. Rese valves Operator opens one valve. Unaffected string will provide Valves to DI Ptnp normally closed necessary flow. Affected string Suction: and are to remain will also provide cooling. Valve close upon ECCS opening has little effect. No D W-75 Operator action. flow is being force (. through the N W-76 Only filters since the letdown flow l

ECCS Sing 13 Failure Analysis - LPI Actuation (Cont'd) Singlo Failure or Component Operator cr Event Ocntrolled By Action Required Error Comment

12. Di Ptap Suction Iso.

V21ves fmm BWST: DN-34 ES-A Valves are nor-Che Valve fails to open. Park lel valve will supply DN-35 ES-B mally open. ES redundant pump string with signal also required flow. forcing them open on LPI Actuation. (Recirculation from Reactor Building Sunp) L3. DH Punp Suction Operator Open valves on Operator does not open valves Operator will get a second alarm Valves from Only BWSr low level (ignores alarm) on LPI low flow at which time he BWST: alam. must restore flow. Ignoring the second alarm would be a DIV-42 second error which is not DW-43 postulated. 1hus, cooling capability not affected. ) ECCS Single Failure Analysis - Core Flood - General, 1. According to B4W Topical Report BAW-10064, April 1973, 'M11tinode Analysis of Core Flooding Line Break for B4W's 2568-MWt Internals Vent Valve Plants," upon a core flooding line break (LOCA), the r-ining core clood tank and one HPI punp string is adequate for proper cooling. 2. Passive failures not considered to occur after IDCA in this system - considered ir. long term cooling only. e 4 8 e --wrr- - v

ECCS Sing 19 Failure Analysis - Gare Flood Single i Failure or Component l Operator. cr Event Odntrolled BV Action Required Error Comments D. Bleed Valves to Valves close Qie valve fails to close or Series valve will close and l Waste Gas System: when RB press. one ES train fails. isolate core flood tank.as req'd > 4 psig. CFV-29 ES-B CFV-16 ES-A t CFV-15 ES-A

6. Sampling Valves:

Valve close when Qie valve fails to close or Series valve will close and RB press. one ES train fails, isolate core flood tank as req's CFV-11 ES-A > 4 psig. CFV-12 ES-A' CFV-42 ES-B 9. Cora Flood Tank Operator Tech Specs 3/4.5 Operator fail's to lock breaker Closure of any of these valves . Outlet Valves to require these and closes a valve. with RC press. > 700 psi result Re:ctor: valves to be in an. alarm. h us, the operato7; open and their must reopen the valves. Ignoring CFV-5 breakers locked this alarm would be a second 1 CFV-6 open and tagged error which is not postulated. before reactor %us both Core Flood Tanks will ' criticality is at all times be in their proper allowed. standby passive condition. 3. IDCA in Core Flood Passive or active failure in loss of one core flood tank and Lins Between Reac. opposite LPI line. all LPI. B4W Topical Report Vessel 4 1st Check 10064 states that adequate Valve cooling will be provided by remaining CFT and one HPI string l l

ECCS Sing 19 Failure hnalysis - Core Flood (Cont'd) Singlo Failure or Component Operator cr Event Contmiled By Action Required Ermr Comments

i. Nitrogen Fill Iso.

Valve close Valve to one CPT fails to 'Ihis valve is normally closed. Vdvas when RB press. close. However, the operator may have > 4 psig been increasing pressure. Operat-CFV-28 ES-A4B ing procedures require that CFV-27 ES-A6B pressure be increased at a rate no faster than 100 psi per 15 min. Since the relief valve is set to relieve at 700 psig, it 4 would take at least 15 min. to increase pressure to this point CFT actuation occurs long before ; this time period is over so the failure of this valve to close has no effect on CFT cooling capability. Watsr Fill Iso. Valve close Valve to one CFT fails to For this non-closure to have any Valv:s: when RB press. close. significance, the II)CA must have > 4 psig had to occur during the time the CFV-25 ES-AGB operator was adding water to the GV-26 ES-A4B CFT through two nomally closed series manual valves in addition l to this valve. Secondly, the water entering the CFT through the 1-inch fill line would have to increase N2 Pressure from a nominal 600 psig to 700 psig where the relief valve would open 'Ihinlly, this pressurization would be a potential problem only in the case of small break when RC pressure decays slowly and prolongs CFT actuation. On large breaks CFT would discharge i long before CFT pressure could increase substantially. Under small break conditions. HPI is already on-line before RC press, drops to 600 psig. Also, CFT -

~~ BCG Singlo Pailure Analysis - Com Flood (Cont'd) Singlo Failure or Component Operator cr Event Captrolled By Action Required Error Coments Water Pill Iso. a Valves: (Cont'd) actuation is not intended to' account for sr.all breaks, but-is primarily for covering the core in case of intermediate to large breaks. In fact, eveg for a break as small as 0.44 fte. CFT actuation would occur before CPT pressure would increase to anywhere near 700 psig. 1hus, this valve failure has no adverse effect on EOCS cooling capabilities. -- -

= ECCS Single Failure Analysis - CFT Water Pill. 9 A. Single Failure: CFT water fill valve does not close when RB press. >>, 4_psig and CFT was being filled at time of LOCA. B. Potential Problem: CFT water fill valve failure has the potential for causing an increase ~in N2 pressure to the point of relief valve actuation before CFT actuation can eccur. C.. Calculation Basis: 1. Relief Valve Set Point: 700 psig 2. The problem is for small breaks only where more time elapses before CFT actuation, thereby allowing more water to enter the CFT and, consequently, cause higher CFT pressure. 3. According to B6W Topical Report 10064, for a break as small as 0.44 ft2, CFT actuation would occur in 140 seconds after LOCA. 4. Design flow for water fill line (1 inch) is 50 gpm. 5. Do not consider that RB pressure increases also, which - applies a back pressure to the relief. valve, requiring, an even higher CFT internal pressure than 715 psia to i relieve - a fact which favors proper CPT actuation. 6. Assume the. CFT pressure increase is adiabatic (This is a conservative assumption an isothermal process). D. Water Flow Into CFT in 140 Seconds 8 50 gpa: 1 50 gym x 140 sec. 60 sed 7 min " 117 gal.== 15.6 ft3 E. CFT Pressure After 140 Seconds, P2-1. For an adiabatic process and an ideal gas: In P + y in V = K (constant) where P = pressure in psia V = volume of gas in ft3 y. Cg, molar heat capacity at constant pressure Cy molar heat capacity at constantnvolume 7 ='1.4 N2..

ECCS Singlo Failure Analysi.s - CFT Water Fill '(cant'd) E. CFT Press'ure Aft'er 140 Seconds. P7 fcon t 'd) 1. (. cont ' d)

Reference:

Physics; Halliday, David and Resnick, Robert; John Wiley.6 Sons; 1966 2. For the CFT: P1 = 615 psia ) ). normal operating conditions 3 Vi = 370 ft )

Thus, in 615 + 1.4 in 370 = K K = 14.7 3.

P2: Pressure after adding 15.6 ft3 of water: In P2 = 14.7-1.4 In 354.4 or P2 = 653 psia or 638 psig F.

== Conclusion:== Water filling due to fill valve failure does not prevent CFT actuation for breaks as small as 0.44 ft2, F e.

QUESTION 2c: FSAR Figure '9-6 shows LPI valves DH-V4A and DH-V4B to be normally closed. To allow low pressure injection subsequent to a CFT line break and.a single active component failure, these valves must be required by Station Technical Specifica-tions to be open, power removed, and breakers locked open. Station. Technical Specifications must also require a periodic test be performed to warn of abnormal leakage of the check valves in the LPI injection lines inside containment and that this test be performed at least annually. These changes provide assurance that abundant core cooling is'available for a CPT line break and further minimize the potential for a LOCA outside containment.

RESPONSE

Valves DH-V4A (FPC Tag. DHV-5 (F)) and DH-V4B (FPC Tag. DRV-6(F)) will be placed in the normally open position. FSAR Figure 9-6. will be revised in Amendment No. 48 to indicate this revision to the Decay Heat Removal System. However, as previously committed to and accepted by the.NRC and ACRS, power must be available to these valves as they are required to be throttled in order to split the decay heat (LPI).. flow. The Low Pressure Injection System is provided with a crossover. line to permit one LPI string flow of 3000 gpm to be split equally, thus providing a minimum of 1500 gpm flow to both core flooding injection nozzles simultaneously should a core ) flooding line or one LPI pump fail. The LPI crossover injec-I tion mode of operation is accomplished by opening the crossover line, provided with a two-way flow element, and remotely adjust-ing the flow through the crossover line to 1500 by throttling the two electric motor operated valves DH-V4A and DH-V4B. (Reference to FSAR page 6-4 subparagraph 6.1.2.1.2 - Am. 32 (10 73) ). Acceptance of this mode of operation by the NRC is further exemplified in the staff's SER on page 6-13 and 6-14, Section 6.3.2 System Design. Therefore, valves DH-V4A and DH-V4B will be placed in the normally open position, but power to these valves must stay connected and the breakers closed. Reactor Coolant System leakage is presently covered in - Section 3/4.4.6 of the Standardized Technical Specifications for Crystal River #3. O . - - _ _ ~

QUESTION 3: It is noted that motor-operated ~ valves from thi BWST are shown norma 11y' closed. It appears that, assuming. sufficient static head were available, the potential for a water hammer'when 3 ECC is injected into a dr? line would be reduced considerably., if these valves were norme11y left open. Please discuss. (Also, confirm that periodic venting of ECCS lines and ECCS pump casings are performed and reference this surveillance. ~ requirement in your station technical specifications.)

RESPONSE

Decay Heat Removal System (Emergency Standby Mode) has been revised to indicate DH-5-A (FPC Tage DHV-34F) and DH-5-B (FPC Tag DHV-35F) to be in the normally open position.

Hence, precluding the possibility of a water hammer due to ECC injection into a dry piping system.

The above revision to the ECC System placing DHV-34F and DHV-35F in the normally open position insures that the system is void of air as the system is now completely filled with water. In addition, the ECCS lines and pumps are vented following-the initial filling of the system from the BWST to remove any trapped air prior to operation. Any leakage in the ECCS is from the system and is replaced from the BWST continuously which precludes the entrance of air into the system. Also, the ECC System is periodically tested as required by the Technical Specifications for Crystal River #3 which additionally insures that no air is contained in the system. For the above reasons, periodic venting of the ECCS lines and pumps is not required for Crystal River #3. .e

QUESTION 4: With regard to the potential for submerged valves, the September 19th letter indicates that there are no valve motors which would be flooded that are required to operate following a LOCA. Provide the level of water assumed after the LOCA. Provide the elevation of the motors for DHR suction valves DHV1 an~d DHV2. Confirm that any.other equipment essential to-ECCS performance which could be submerged would not be adversely affected. .... e

RESPONSE

The'poc4 LOCA flood level is 99.85. The' elevations of the DHR suction valves DHV-1, 2-6 3 (FPC Tag Nos. DHV-3F, 4F 6 41F) are as follows: DHV-1 (DHV-3F) 115'-0" DHV-2 (DHV-4F) 106'-0" DHV-3 (DHV-41F) 106'-0" No other equipment essential to ECCS performance would be submerge.dr

• e o

4 v

QUESTION 4: With regard to the potential for submerged valves, the September 19th letter indicates that there are no valve motors which would be flooded that are required to operate following a LOCA. Provide the level of water assumed after the LOCA. Provide the elevation of the motors for DHR suction valves DHV1 and DHV2. Confirm that any other equipment essential to ECCS performance which could be submerged would not be adversely affceted.

RESPONSE

The post LOCA flood level is 99.85. The elevations of the DHR suction valves DHV-1, 2 4 3 (FPC Tag Nos. DHV-3F, 4F 4 41F) are as follows: DHV-1 (DHV-3F) 115'-0" DHV-2 (DHV-4F) 106'-0" DHV-3 (DHV-41F) 106'-0" No other equipment essential to ECCS performance would be submerged. e i s m - _. _,,,,., _. -. _ _ - -.,,, -, ~ _. - - -

QUESTION 5. With regard to the partial loop analysis (Reference to BAW-101C;): a. Technical Specifications will prohibit 2-pump operation unless an analysis is provided to support his mode of operation. Compare a break in the inactive cold leg to a break in the active cold leg. RESPONSE. ~. Based on the arguments presented below, conformance to the ECCS acceptance criteria given in 10 CFR 50.46 is assured for 2-pump operation. With 2 pumps operating, one in each loop, the maximum power level will be 51% of full power which includes 2% for un-certainty. The system flow rato is reduced to 50% of normal 4-pump operation at steady state conditions. The idle pumps in each loop are locked in position because flow is reversed in each of the inactive cold legs. About 18.8% of the RC flow from the downcomer plenum is directed back in each inactive cold leg. The core flow for a break in the inactive pump with 2 pumps operating should be similar to a break in the active cold leg-inactive loop with 3 pumps operating. During the LOCA transient, the positive driving force for both breaks is with 2 pumps, and therefore the core flow should be approximately the same. The reflooding rates for the 2-pump case should be greater than the 3-pump case because the core power is lower, 51% versus 77% of full power rating, thus a lower cladding temperature rise after the E0B would be expected for 2 pumps operating. A break at the pump discharge of either one of the active cold legs will cause a loss in positive flow during the first half of the transient compared to the above case. The transition from positive to negative flow should occur earlier. The negative flow would be substantially increased due to the decrease from 2 to 1 active pumps trying to force positive flow into the core region. The high negative flow rate through the core during the blowdown phase should provide good core cooling and remove a significant amount of stored energy in the fuel. Thus the cladding temperature during this phase of the LOCA should be maintained at a relatively low temperature. The re-flooding phase should have the same improvement in clad tempera-ture as described for the previous 2-pump case, i.e., a lower cladding temperature rise after the E0B would be expected. 9 ~ Therefore, the~ maximum cladding temperature for 2-pump operation should be approximately equal to or less than that calculated for 3-pump operation. Since the calculated peak cladding temperature for a LOCA that occurs during 5-pump operation gives a large margin (434F) relative to the 2200F limit, 2-pump o?eration will easily comply to the acceptance criteria for tie ECCS set forth in 10 CFR 50.46. 9 9 t 2

QUESTION 5. b. Provide issurance (other than historic extrapolation)- that the' PCT versus Break Size curve in BAW-10103 would not be significantly altered by either mode of partial loop operation. RESPONSE. .- 1 Core' flow for a break in the inactive loop.-active cold leg. and'ir. active loop. inactive cold leg is similar to the core flow'with pumps tripped and powered for 4-pump operation respectively, as can be seen by comparing Figures A-2 and A-4 of the partial loop analysis to Figure 5-7 of BAW-10103. Therefore,. core flow for smaller breaks during partial loop operation would be similar to that shown in Section 6 of BAW-10103. With similar flows, the PCT versus Break size curve should exhibit the same trend, i.e., decreasing P.CT with-break size. Since the PCT for the p~artial loop analysis is 313F less than that given for the worst break in BAW-10103,-. smaller breaks will exhibit large margins of safety relative to the 2200F criteria. e 4 e 9

L a QUESTION 5. Submit the LOCA parameters of ~inte' est identified in the c. r " Minimum Requirements for~ECCS Break Spectrum Submittals" dated April 25, 1975. RESPONSE. The following are additional LOCA parameters of interest ~ for the B4W category 1 partial loop LOCA analysis. 3-Pumps, Break at Active Pump Inloop With Idle Pump (CRAFT Run PP 102 (Yl)) Figure 1. Reactor Vessel Pressure for 8.55 ft2 DE Break at Pump Discharge During Partial Loop Operation, CD = 1.0 Figure 2. Core Water Level for 8.55 ft2 DE Break at Pump Discharge During Partial Loop Operation, Cp = 1.0 Figure 3. Downcomer Water Lovel for 8.55 ft2 DE Break at Pump Discharge Partial Loop Operation, CD

• 1-)

Figure 4. Total Power for 8.55 ft2 DE Break at Pump Discharge During Partial Loop Operation, CD = 1.0 Figure 5. Containment Pressure for 8.55 ft2 DE Break at Pump Discharge during Partial Loop Operation, CD = 1.0 3-Pumps, Break at Idle Pump (CRAFT Run PP 101 (IB)) Figure 6. Reactor Vessel' Pressure for.8.55 ft2 DE Break at Pump Discharge During Partial Loop Operation, Cp = 1.0 Figure 7. Core Water Level for 8.55 ft2 DE Break at Pump Discharge During Partial Loop Operation, CD = 1.0 Figure 8. Downcomer Water Level for 8.55 ft2 DE Break at Pump Discharge During Partial Loop Operation, CD = 1.0 Figure 9. Total Power for 8.55 ft2 DE Break at Pump Discharge During Partial Loop Operation, C = 1.0 D l...

Qaput:r Data for the Figures Fig. No. Version Name Version Date Run.Name Rm Date 1 CRAFP 2, Version 5 PP 4/17/75 PP102(Y1) 07/25/75 2,- RELF00D 2, No Ioop 12/20/74 PR102(2I) 07/28/75 Version 2 3 REFIDOD 2, No loop 12/20/74 PR102(2I) 07/28/75 Version 2 4-GAFT 2,_ Version 5 PP 4/17/75 PP102(Y1). 07/25/75 5-CNTEMPT, Version 15 11/15/74 PC100(FR).. 07/11/75 6 CRAFT 2, Version SPP 4/17/75 PP101(1B). 07/15/75 7' REFID0D 2, No Ioop 12/20/74 PR101(NJ) 07/15/75 Version 2 8 REFIDOD 2, No Inop 12/20/74 PR101(NJ) 07/15/.75 ~ Version 2 9 CRAFT 2, Version SPP 4/17/75 PP101(1B) 07/15/75 Other Codes Used (Figures provided in original report) Version Name Version Date Run Name Rm Ihte 'INETA 1B, Version 6F 1/23/75 FIl01(IJ) 07/18/75 'INETA IB, Version 6F 1/23/75 Fr1A2(HT) 07/30/75 .= O'

o S e.oco -

5.cc3 -

2 Figure 1. Reactor vessel ?ressure for 8.55 f t DE Break at Pump Discharge During Partial Loop Operatiop, ~ C = 1.0. D 30 O - a e x = ?$.@ - ? = 20.M ' i M I to a-15 0':3 - taJ 4 OC e D 6

• i.

nn i Es1 M 10 023 - n. 8 q Ie e 5.Geo .e a l e li.M N.M N@ SkW.. I d.m d.Ico 7 900 38.808 li M 35'.m i

o. goo Tine,. sac j

4 5 e = -- m

o , j- ~ 14.o15 ' i _. 2 Figure 2. Core Water Level for 8.55 ft DE Brqak'at Pump Discharge During Partial Idop'Operettee, 1" c = 1.0. a no.oco.. t s.o:n_ LJ. D i M f e.oco. M a. w u se

e. ace.

O t u 1 -e g,eaa, t e o.an. 6 -l o.ons

3. w.

s.eso . 7.$os,',,te q ' y.nep. . it.quigi tv.sco'. nov p.sge,. I .g ~ ~ Time aftar 208.ses (u..l0.). ~ p

/ e 3

e.cz 2

Figure 3. Downcomer Water Level for 8.55 ft DE Break I' 8 - at" Pump Discharge During Partial Loop Operation, C,= 1.0.

om.

6 a 14 Ifb.GC3. G s r cc I:.000. a: w E o M s.cm_ ~ es es 4mm. o.oca. i I V I 1 I T I I "".on-sm s,000 7m 50 000 13 2 o 85 000 ' t?.m 88900 82 488 8 p 3 TDie after 503,ses (x10 ) 3 n n 1

_____-____-_________________________g. ~ t.633 1'*' 2 DEBreakatth Figure 4. Total Fever for 8.55 ft Discharge During Partial Loop Operation. C = 1.0. D 1 2tc - I e t.033 - 0 3 i g, 42-l l a . a: e-O p. .g. e .g =- o n- .e* ~

h.-

e U* ' ' '[ ..~~g.,! . i: ' ',. 7 1: t I v.:=s ., j ~ji: ;7,s. <~,'U 1 ;.. I.T r / 'q 9 333 ' E.400.. 8 888. 7.000 T,- se,aos,

33. sos.

33.ggg, 3g, 9 *, g.ggg' gs, egg ._ gag T.Itt, see. + ~ 3 4 o I n ' o

.u .' s e .ma.w l m.s 2 ~ Figure 5. Containment Pressure for 8.55 f t DE Break j at Pump Discharge During Partial Loop Operation .m.m C = 1.0 i. D T...[ .82.1. *.* lI 1 J. i w .m z g l b .asA.u w r ..g g

• en.T 8

o 9 o 2.= y z l . >= W .=e .u.c g C a N.,. I i u .eu.s / / .va.e e e / ..e.. ,/ / 6 gg.h ._ L li j 16* 10 10' 10' 10 10.8 8 i ilHE ffif.R RUPTURE.SF.C. 0, g p

..) n e.co: as.coe - 's 2 Figure 6. Reactor vessel Pressure for 8.55 (t DR Break at Pump Discharge During Partial Loop Ope'ratian, ,{ y,coe. C = 1.0. y D o N w

• 0.or -

4 w SuBD C Q. 15 00:- e h.s 2 l =a ~1 $ 10 003 - i c 52-

• f I

.e = a e daca si.uo A' m doce dW. l.saa i.m i.oon W,noe o.aca TIME,sec- -i. 4 r s .i t l. i e e I O s 9 e 9 e. '. O

o ~. l l I r 1 ~ 14.015 tr.om. Core Water Level for-8.55 f t2,,,,,,g,, Figure 7. Pump Discharge During Partial Loop'Operatios,. ic. con-C D" s.cm_ w a y = us s s.cm_ mg .x s

i W

e.as:_ a a s.ex. r t r ~ e c r som : IJ%. 1 h.oag 7.Scti '10 000 12.500 15400 57993

1. 9@ '

g.gog g.Soj 3 o. Time after 505 sec (x10 ) e.' ^ 0 a

o ~ a l l ? ~ \\ 6 2s.000 ~ Figure 8. DowncomerWaterLevelfor8.55itfDERyegh 24.m-at Pump Discharge During Partial Leop.6pera41on, C *= 1.0. ~ D ro.txu_ r. w .e u is.oz. u r J a: ss.cz_ s .ar: a r.o sd 2 s.coc_ n, T o 44m_ o 3 esv- \\ . =.., - a :. -- e 0,.923 8y $,.' SM .':.,.I,1%.' p.033 'J$, 89%..J,. tty. 45m '

• Q.%
• 9 e

g ..Tilme af ter E08.see (x10 ) ~-, e. a b e' e 0 d

} o s.sz s.m - 2 Figure 9. Total Fever for 8,55 ft DE Break at Femmy Discharge During Partial Loop Operation, C.= 1.0. D i.e 2 A=' b,,; ce e I 1 S l O l l .m. l l .m : ,s i 1 r. c.cn d.m yj,q p',g , g,p ^ 8M 2 805 3 700 - 7.800 lo eco p,4'.., yg TiflE,sao I ^*

J.

a.. m. . u 4

? .~ QUESTION.5. ~ ~ d. The analysis states that the peak linear heat rate for the hot bundle is the maximum kw/ft LOCA limit as shown in Figure 2-2 of BAW-10103 at the six foot elevation for this made of operation. Clarify the effect of such an assumption on Station Technical Specifications. If this assumption would allow 3-pump operation at this kw/ft, discuss the consequences of the abnormal events in Chapter 15.0 should they occur during., three (or two) pump operation. RESPONSE. The Station Technical Specifications account for the maximua kw/ft limit for partial pump operation in the imbalance safety.. limits reported in Section 2.1 of the Tech Spec. The abnormal events analyzed in Chapter 14.0, are initiated from nominal' opers ;ing conditions to obtain the most severe core average results. For those accidents in which fuel clad failure is determined, hot channel DNB calculations are performed to account for the maximum peaking condition in the hot pin. The four pump, rated power cases presented in Chapter 14.0 result in the most severe consequences. In addition, to initiate a transient from other than design peaking conditions results' in consequences (DNBR) equivalent or less severe, as design. peaking represents the limiting case. I i 1

i. 3 :- ,\\..t ( QUESTION 6: Provide your schedule for submitting revisions to the proposed.-- Technical Specifications affected by the LOCA analysis.

RESPONSE

As indicated by.the above responses, there are no proposed revisions to the Technical Specifications for Crystal River #3 at the present time, as a result of the LOCA analysis. l f i _ - _. _ _,}}