ML19329F916

Forwards Response to NRC 800310 Request for Info Re Auxiliary Feedwater Sys Design Basis Info & Pump Flow Verification,For Unit 2.Unit 1 Info Will Follow Shortly

ML19329F916
Person / Time
Site:	North Anna
Issue date:	07/10/1980
From:	Sylvia B VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	Harold Denton, Youngblood B Office of Nuclear Reactor Regulation
References
592, NUDOCS 8007110359
Download: ML19329F916 (15)

Text

4 o,. *,.o Vruoxxtr ELucrute Axn Powen CoxeAxv Hrczrxoxo,Vshora sA unum July 10, 1980 Mr. Harold R. Denton, Director Serial No. : 532 Office of Nuclear Reactor Regulation N0/LEN/jmj Attentlan:

Mr. B. Joe Youngblood, Chief Docket No. 50-339 Licensing Branch 1 License No. NPF-7 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C.

20555

Dear Mr. Denton:

NORTH ANNA UNIT 2 AUXILIARY FEEDWATER SYSTQi, REQUIREMENTS We have reviewed your letter of March 10, 1980 that requests information con-cerning auxiliary feedwater system design basis information and pump flow verification. Our response to these items is provided in Attachment 1.

We believe that the attached response adequately addresses the NRC concerns for North Anna Unit 2 and that a coccitment to provide addit-ional information at a later date is not necessary. Our response for North Anna Unit 1 will follow shortly. Should you have any questions or require additional informa-tion, please contact us.

Very truly yours,

hk B. R. Sylvia Manager - Nuclear

~

Operations and Maintenance Attachment

- 8007110 3d 0

(

ATTACIBtENT 1

..1 Ques:1 c n *_

dentify the alant transient and accident conditions considered in a.

establishing AFWS ficw recuirements, including the following events:

1)

Loss of Main Feed (LMFW)

2) LMFW w/ loss of offsite AC ocwer 3)

LMFW w/ loss of onsite and offsite AC power 4)

?lant ecoidown 5)

Turbine trip with and without bypass

5) Main steam isolatien valve closure

7) Main feed line break S) Main steam line break 9)

Small break LOCA

10) Other' transient or acciden: conditions not listed above.

b.

Describe the plant protection ac:ectance critaria and corresponding technical bases used for each initiating event identified above.

The acceptance criteria should address plant limits such as:

1) Maximum RCS pressure (PCRV or safety valve actuatien) 2)

Fuel temperature or damage limits (DNS, PCT, maximum fuel central temperature)

3) RCS cooling rate limit to avoid excessive ecolant shrinkage I

a) Minimum steam generator level to assure sufficient steam gen-erator neat transfer surface to remove decay heat and/or c:01 dcwn the primary system.

Resconse to 1.a The ' Auxiliary Feedwater System serves as a backup system for supplying

~

feedwater to the secondary side of the steam generatcrs at times when the feedwater system is not available, thereby maintaining the heat sink cacacilities of the steam genert:or.

As an Engineered Safeguards Sys-tem, the Auxiliary Feedwater System is directly relied upon to prevent c:re damage and system overpressurizaticn in the event of transients such as a loss of normal feedwa er er a secondary system pip,e rupture,.

and to provide a means for p1an: ::ald:wn foilcwing any plant transient.

Following 'a reactor. trip, deccy heat is. di.ssipated by evaporating' water in the steam generators and venting the generated' steam either to the condensers through the steam dump or to the atmcsphere through the steam generat:r safety valves or the power-operated relief valves.

Steam generat:r water inventcry must ce maintained at a level sufficient to ensure adecuate heat transfer and centinuation of the decay heat removal process.

The water level is maintained under these circumstances by the

. Auxiliary Feedwater System which delivers an emergency water supply to the : steam generators.

The Auxiliary Feedwater System must be capable Of func icning.fcr extended periods, allowing time either to restore normal feecwater flew or to proceed with an cederly cooldcwn cf the plant t:

. o ~.e

. the reactor coolant temaer1ture where the Residual Heat Removal System can assume the burten of decay heat removal.

The Auxiliary Feedwater System ficw and the emergency water suoply capacity must be sufficient to remove core :ecay heat, reactcr. coolant pump heat, and sensible heat during the plant cccidown.

The Auxiliary Feedwater System can also be used to maintain the steam generator water levels above the tubes fol-lowing a LOCA.

In the latter function, the water head in the steam generators serves as a carrier to prevent leakace of fission products from the Reactor Ccolant System into the secondiry plant.

DESIGN CONDITIONS The reactor piant c:nditions which impose safety-relatec cerformance requirements on the design of the Auxiliary Feedwater System are as follows for the North Anna Units 1 and 2 plants.

Loss of Main Feedwater Transient Loss of main feedwater with offsite power available Station blackout (i.e., loss of main feedwater without offsite oower available)

Secondary System Pipe Ruptures Feedline rupture Steamline rupture Loss of all AC Pcuer Loss of Coolant Ac:ident (LOCA)

Cooldown Loss of Main Feedwater Transients The design. loss of main feedwater transients are those caused by:

Interruptiens of the Main Feedwater System fiow due to a malfunction in the feedwater or condensate system j

Loss of off site power or biackout with the consequential shutdcwn of i

the system. atmos, auxiliaries, and controls j

Loss of main feedwater transients are characterized by a rapid reduction in steam generat:r water levels which results in a.reacter trip, a turbine trio, and auxiliary feedwater actuation by the protection system logic.

Following reactor trip frem a high initial cower level, the power quickly

-falls to decay heat levels.

The water levels continue to decrease, progressively uncovering the steam generator tubes as decay heat is transferred and tischarged in the form of stern either through the steam duma valves to the cendenser or through the stesa generator safety or.

l Ocwer-coerated relief valvas to the atmosphere.

The reactcr coolant

o

. temperature increases as the residual heat in excess of that dissipated through the steam generators is absorbed. With increased temperature, the volume of reactor coolant expands and begins filling the pressur-izar. Without the addition of sufficient auxiliary feedwater, further expansion will result in water being discharged through the pressurizer safety and/or relief valves.

If the temperature rise and the resulting volumetric expansion of the primary coolant are permitted to continue, then (1) pressurizer safety valve capacities may be exceeded causing overpressurization of the Reactor Coolant System and/or (2) the continu-ing loss of fluid from the primary coolant system may result in bulk boiling in the Reactor Coolant System and eventually in core uncovering, loss of natural circulation, and core damage.

If such a situation. were ever to occur, the Emergency Core Cooling System would be ineffectual because the primary coolant system pressure exceeds the shutoff head of the safety injecticn pumps, the nitrogen over-pressure in the accumula-tor tanks, and-the design pressure of the Residual Heat Removal Loop.

Hence, the timely introduction of sufficient auxiliary feedwater is necessary to arrest the decrease in the steam generator water levels, to reverse the rise in reactor coolant temperature, to prevent the pressur-izer from filling to a water solid condition, and eventually to estab-lish stable hot standby conditions.

Subsequently, a decision may be made to. proceed with plant cooldown if the problem cannot be satisf ac-torily corrected.

The blackout transient differs from a simple loss of main feedwater in that emergency power sources must be relied upon to operate vital equip-ment.

The loss of power to the electric driven condenser circulating water pumps results in a loss of condenser vacuum and condenser dump valves. Hence, steam formed by decay heat is relieved through the steam generator safety valves or the power-operated relief valves.

The calcu-lated transient is similar for both the loss of main feedwater and the blackout, except that reactor ccolant pump heat input is not a censider-ation in the blackout transient following loss of power to the reactor coolant pump bus.

Secondary System pioe Ruotures The feedwater line rupture accident not only results in the loss of feedwater flow to the steam generators but also. results in the complete blowdown of one steam generator within a short time if the rupture should occur downstream of the last nonreturn valve in the main or auxiliary feedwater piping to an individual steam generator.

Another significant result of a feedline rupture may be the spilling of aux-iliary feedwater out the break as a consequence of the fact that the suxiliary feedwater branch line may be connected to the main feedwater line in the region of the postulated break.

The system design must, allcw for terminating, limiting, or minimizing that fraction of auxiliary feedwater flow which is delivered to a faulted loop or spilled through a break and to ensure that sufficient flow will be delivered to the remaining effective steam generator (s). The concerns are similar for the main feedwater line rupture as those explained for the loss of main feedwater transients.

57J1A

4-Main steamline rupture accident conditions are characterized initially by-plant cooldown and, for breaks inside containment, by increasing centainment pressure and temperature. Auxiliary feedwater is not needed during the early phase of the transient but flow to the faulted loop will contribute to the release of mass and energy to containment.

Thus, steamline rupture conditions establish the upper limit on auxiliary feedwater flow delivered to a f aulted loop.

Eventually, however, the-Reactor Coolant System will heat up again and auxiliary feedwater flow will be required to be delivered to the unfaulted loop, but at somewhat lower ratas than for the loss of feedwater transients described pre-viously.

Provisions must be made in the design of the Auxiliary Feed-water' System to allow limitation, control, or termination of the auxil-iary feedwater flow to the f aulted loop as necessary in order to prevent containment overpressurization following a steamline break inside con-tainment, and to ensure the minimu'm flow to the remaining unf aulted loops..

Loss of All AC Power The loss of all AC power is postulated as resulting frcm accident con-ditions wherein not only onsite and offsite AC power is lost but also AC emergency power is lost as an t sumed common made failure.

Sattery power for operation of protection circuits is assumed available.

The impact on the Auxiliary Feedwater System is the necessity for providing both an auxiliary feedwater pump power and control source which are not dependent on AC power and which are capable of maintaining the plant a.t hot shutdown until AC power is restored.

Loss-of-Coolant Accident (LOCA)

The loss of coolant accidents do not impose on the auxiliary feedwater i

system any flow requirements in addition to those required by the other accidents addressed in this response.

The following description of the small LOCA is provided here for the sake of completeness to explain the role of the auxiliary feedwater system in this transient.

Small LOCA's are characterized by relatively slow rates of decrease in reactor coolant system pressure and liquid volume.

The principal con -

tribution frem the Auxiliary Feedwater System following such small LOCAs is basically the same as the system's function during hot shutdown or following spurious safety. injection signal which trips the reactor.

Maintaining a water level inventory in the secondary side of the steam generators provides a heat sink for removing decay heat and establishes the capability for providing a buoyancy head for natural circulation.

The auxiliary feedwater system may be utilized to assist in a system cooldown and depressurization following a small LOCA while bringing the

. reactor to a cold shutdown condition.

l 5741A' e

,7 e

e 5

C ::dcwn The : oldewn function performed by the Auxiliary Feedwater Systen is a partial one since the reacter coolan system is reduced fr:m normal zero loac temoeratures to a hot leg temoerature of accroximately 350cF.

The latter 's the maximum temperature recc. mended for placing the Resi-dual Heat Removal System (RHRS) into service.

The RHR system ccmpletes the cooldcwn to cold shutdewn conditions.

Cooldown may be required following excected transients, following an accicent such as a main feedline break, c; during a normal cocidown prior to refueling or performing reactor plant maintenance.

If the reactor is tripped following extended operatien at rated power level, the AFWS is capable of delivering sufficient AFW to remove decay heat and reactor coolant pumo (RCP) heet following reactor trio while main-i taining the s eam generator (SG) water level.

Fo11cwing transients or accidents, the recc. mended cocidewn ate is consistent with expected needs and at the same time does not irpose additional requirements on the cacacities of the auxiliary feedwater cum:s, considering a single failure.

In any event, the process censists' of being able to dissipate plant sensible heat in addition to the decay heat produced by the reac-tor core, i

I 9

4 9

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-6

.Rescense to 1.o Table 13-1 summarizes she criteria which are the general design bases ~

for each event, discussed in the rescanse to Question 1.a, above.

Specific assumations used in the analyses to verify that the design bases are met are discussed in resconse to Question 2.

The-primary function of the Auxiliary Feedwater System is to provide sufficient heat removal cacacility for heatuo accidents folicwing reac-tor trip to remove the decay heat generated by the core and prevent system overoressurizatien.

Other plant protection systems are designed to meet short tcm or ore-trip fuel failure criteria.

The effects of excessive c:olant shrinkage are bounded by the analysis of the rupture of a main steam pipe transient.

The maximum flow requirements deter-mined by other bases are incorporated into this analys,is, resulting in no additional flow requirements, a

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-e-Question 2

' Describe the analyses and assumotions and corresconding technical justi-fic2 tion used with plant condition considered in 1.a above including:

Maximum reactor power (including instrument error ailowance) at the a.

time of the initiating transient or accident.

.b.

Time delay fecm initiating event to reactor trip.

c.

Plar.t parameter (s) which initiates AFWS flow and time delay between initiating event and introduction of AP45 flow into steam genera-tor (s).

d.

Minimum steam generator water level when initiating event occurs.

e.

Initial steam generator water inventory and depletion rate before and af ter APAS flow comences -- identify reactor decay heat rate

used, f.

Maximum pressure at which steam is released fract steam generator (s) and against which the APA pump must develop sufficient head.

g.

Minimum number of steam generators that must receive AF'd flow; e.g.,

1 out of 2? 2 out of 47 h.

RC flow condition -- continued operation of RC pumas or natural

?

circulation.

1

i. Maximum APA inlet temperature.

j. ' Following a postulated steam or feed line break, time delay assumed to isolate break and direct AF4 ficw to intact steam generator (s).

AFW pump ficw caoacity allcwance to acccmodate the time delay and maintain minimum steam generator water level.

Also identify crecit taken for primary system heat removel due to blowdcwn.

Volume and maxiaum temcerature of water in main feed lines between steam generator (s) and AF'45 ::nnection to main feed line.

i. Operating condition of steam generator normal blowdown failowing initiating event.

?rimary and secondary sys' em water and metal sensible heat used for t

m.

cooldown and AF4 flow sizing.

n.

Time at hot standby and time to cooldown RCS to RHR system cg in temperature to size APA water source inventory.

57alA

. Resconse to 2 Analyses have_ been aerformed for the limiting transients which define the AFWS :erformance requirements.

These analyses have been provided for review and have been approved in the Aoplicant's F3AR.

Specifi-cally, they include:

Loss of Main Feedwater (Station Blackout)

Rupture of a Main Feedwater Pioe Ruoture of a Main Steam Pipe Inside Containment In adv.ition to the above analyses, calculations have been performed specifically for the Nor:n Anna Units 1 and 2 plants to determine the

. plant c oidewn flow (s:Orage capacity) requirements.

The Loss of All AC Power is evaluated via a :cmoarison to the transi en results of a Blackout, assuming an available auxiliary pump having a diverse (non-AC) power sucoly.

The LOCA analysis, as ciscussed in response 1.b, inco -

porates the system ficws recuirements as defined by other tr?.nsients, and therefore is not oerformed for the cur ose of scecifying AFWS flow requirements.

Each of the analyses listed acove are explained in further detail in the folicwing sections of this response.

Loss of Main reedwater (Blackouti A loss of feedwater, assuming a loss of power to the reacter coolant pumos, was aerformed i, FSAR Section 15.2.3 for the purpose of showing that for a station blackout transient, the peak RCS pressure remains below the criterion for Condition II transients and no fuel f ailures occur (refer to Table 13-1).

Table 2-1 summari:es the assumations used in this analysis.

The transient analysis begins at the time of reactor trio.

This' can be done because the tris accurs en a steam generator level signal, nence the core ocwer, temseratures and steam generator level at time of reactor trip do no deceno en the event secuence prior to trip.

Althougn the time from the less of feedwater until the reactor trio occurs cannot be deternined from this analysis, this delay is ex:ec:ec to be 20-30 seconds.

The analysi.s assures that the plant is initially coerating a: 102% (calorime:ri: error) of the Engineerec Safeguards cesign (550) rating sncwn en :ne table, a very conservative assumption in defining decay heat ant stored energy in the RCS.

The

-reac::r is assumed to be tripoed on icw-low steam generator level, allcwing for level' uncertainty.

The 5AR shcws that :nere is a con-siceracle margin with respect to filling the pressurizer.

Ru :ure of Main Feedwater Pice The : uble ended rupture of a main feecwater cipe dcwnstream of the main feedwater line check valve is analy:ed in FSAR Section 15.2.2.2 (see.

also 515.13).

Table 2-1 summarizes the assumotions usa'd in'this an alys i s.

Reac:Or tric is assumed Oc :ccur when the unaffected steam

enera

Ors ' ara at :he icw level set:oint (adjus:eo f ar errors) and the f aulted locp is assumed to be at the low-icw level tric setpoint.

o11 s

nserva-ive assumotion maximi:es the stored heat price tc reactor tric ind minimizes tne aoility. cf the staam generator tc rencve heat fecm Ne RCS following reactor tric cue to 1 conservatively smal' total steam

-9 generat:r i: vert:ry.

As in the loss of normal feedwater analysis, the ini-i11 :cwer cating was assumed to be 102". of the ESD rating.

Auxi-liary f eedwater ficw of 3aC g;m aas assumed to ce delivered to one non-faulted steam cenerator one tinute after reactor trio.

Thirty minutes afur the ac:fdent, the auxiliary feedwater rate is increased to 350 gem divided between two steam generat:rs, as a result of the realignment of valves by the operator.

The criteria listed in Table 13-1 are met.

This analysis estaolishes requirements for layout to preclude indefinite loss of auxiliary feedwater to the postulated break, and establishes train asscciation requirements for equipment so that the AFWS can deliver the ninimum flow required in one minute following op.erator actions assuming the worst single failure.

Ruoture of a Main Steam Pios Inside Containment Because the steamline break transient is a cooldown, the AFWS is not needed to re.TcVe heat in the short term.

Furthermore, addition of excessive auxiliary feedwater to the f aulted steam generator will affect tne peak c:ntainment pressure following a steamline break inside con-tainment.

This transient is cerformed at three power levels for several break si:es. Auxiliary feedwater is assumed to be initiated at the time of main feedwiter isolation, i1 dependent of system actuation signals.

The maximum ficw is used for this analysis, considering a pump runcut.

Table 2-1 sumarizes the assumotions used in this analysis.

At 30 minutes af ter the break, it is assumed that the operator has isolated b

the AFWS fr:m the f aulted steam generator which subsequently blows down A

to ambient oressure.

The criteria stated in Taole 13-1 are met.

This transient estaclishes the maximum allcwable auxiliary feedwater flow rate to a single f aulted steam generator assuming all pumps opera-ting, establisnes the basis for runcut crotection, if needed, and establishes 1ayout requiremer.ts so that the flow requirements may be met considering the worst single failure.

P1 ant Cooldcwn Maximum and minimum ficw recuirements fr:m the previously discussed transients meet the ficw requirements of plant cooldown.

This coera-tien, hcwever, defines the basis for tankage siae, based en the required c:cidewn duration, maximum decay heat inout and maximum stored heat in the system.

As crevicusly discussed in rescanse 1A, the auxiliary feed-water system cartially cools the system to the point where the RHRS may c:malete the cocidown, i.e., 3500F in the RCS.

Table 2-1 shows the assumoti:ns used to determine the cooldown heat cacacity of the auxil-iary feecwater system.

The cooldown is assumed to commence at the maximum rated power, and maximum trio :el sys and decay heat source. terms are assumed when the reactor is tri: ped.

Primary metal, primary water, secondary system metal anc 'sec nciry system water are all included in-the. stored heat to be removec by d e ?FWS.

See Taoie 2-2 for the items constituting the sensible heat st red in the NS53.

his coerati:n is analyzed to estaolish minimum tank size requirements Sr auxiliary fee:wner fluid scur:e which are ncrmally aligned.

TABLE 2-1 SUMMAkY OF ASSUMPTIONS USED IN AFWS DESICN VERIFICATION ANALYSE 3 Loss of FecJwatar Main Steamline Break.

Transient (station blackout)

Cooldown E11n Faedlina Break (containment)

Max reactor power -

102% of ESD rating 2963 MWt 102% of ESD rating 0% (of rated) - vorst (102% of 2910 MWt)

(102% of 2910 MWt) case (% of 2775 MWt) b.

Tino delay from 2 see (delay after 2 sec

,2 sec 0 seconds event to Rx trip trip) c.

AFWS actuation signal /

lo-lo SG 1evel NA low-lov SG 1evel Feodwater isolaticn actuation signal:

time delay for AFWS flow 1 einute 1 minuto Safety Injection Signal 0 see (no dalay);

Flow to SG: 17 see after MSLB (which is tima of FW isol.7 tion) d.

SC water level at (lo-lo SC 1cyc1)

NA (lo SG 1evel.+ steam-Sama as initial level beforo event tima of reactor trip 0% NR span feed mismatch) 2 0 20% NR span 1 0 tube sheet e.

Initial SC inventory 74.600 lbm/SC (at 134.500 lbc/SG 94.100 lb=/SG 151.000 Itm/GG trip)

G 525.2*F Rate of change before Sco FSAR Figure 15.2.31 N/A turnaround greater FSAR Table 6.2-11

& at ter AFWS actuation than 2000 sac.

decay heat See FSAR Figure 15.1-6 See FSAR Figure 15.1-6 See FSAR Figure 0.2-1 f.

AFV pucp design 1133 psia 1133 psia 1133 psig 1025 psig pressurc 3

Minimum # of S0s 2 of 3 N/A 1 of 3 (1 min after N/A which =est reccive trip) 2 of 3 (cporator AF4 flow cetion af ter 30 minutes) h.

RC pump status Tripped G reactor trip Tripped All operating Tripped 1.

Msximu= AF'J 120*F 100'F 120*F 120*F Tamparature j.

Operatcr cetion nono N/A 10 min.

30 min, 3

3 3

3 k.

MFV parga volumc/ temp.

150.6 ft /437.6'F 100 ft /

214 ft /433.3'F 218 ft /441*F 434.3*F 1.

. Nor=al blowdawn.

none assumed none ascumed none assumed nono assumed

=.

Sensible heat sea cooldown Tabla 2-2 see cooldown sea cooldown n.

Timo at standby /timo

-2 hr/4 hr 2 hr/4 f.r 2 hr/4 hr N/A to :coldown ta RER o.

AFW flow rate 680 CPM - constant variablo 340 epm - 1 min af ter 900 gpm to broken SG trip - 630 gpm 350 Epm to each intact S/C (af ter 30 minutes)

(zax. require =ent)

TABLE 2-2 Summary of Sensible Heat Sources Primary Water Sources (initially at rated power temoerature and inventory)

- RCS fluid

- Pressurizer fluid (liquid and vacor)

Primary Me'tal Sources (initially at rated power temperature)

- Reactor coolant piping, cumps and reactor vessel

- Pressurizer

- Steam generator tube metal and tube sheet

- Steam generator metal below tube sheet

- Reac.or vessel internals Secondary Water Sources (, initially at rated power temperature and inventory)

- Steam generator fluid (liquid and vapor)

- Main feedwater purge fluid between steam generator and AFWS piping.

Secondary Metal Sources _(initially at rated power temperature)

- All steam gen.erator metal above tube sheet, excluding tubes.

9

+

t

Question'3 Verity that the AFW pumps in your plant will supply the necessary flow to the steam generator (s) as determined by items I and 2 above considering a single failure.

Identify the margin in sizing the pump flow to allow for pump recir-culation flow, seal leakage and pump wear.

Response to 3 Based upon review of the flow requirements of Table 2-1, the main feedline break was selected as the mos,t conservative accident in

e. valuating the AFW pump flow rates based'on flow at pressure.

The parameters used are as follows:

1.

Steam Generator Pressure 1133 psia 2.

AFW Temperature 120*F 3.

Flow Rate to any Steam Generator (min) 340 gpa The motor driven AW pumps were originally specified to deliver 350 gpa at 1216 psi.

The vendor design point included recircula t ion flow and became 370 gpm at 1214 psi.

The turbine driven pump was specified to deliver 700 gpm at 1215 psi.

The vendor design point included recirculation flow and became 735 gpm at 1214 psi.

Certified pump performance curves from Intersoll-Rand verify that these design points were met.

Recent system pressure drop calculations using 350 gpa as a design flow rate have been compared with the vendor certified pump test curves.

The results indicate that a pump Nad margin exists at the flow required by W (340 gpm),

for all three North Anna Unit 2 auxiliary feed pumps.

The margin of pump head or flow assigned for seal leakage and pump wear is essentially zero.

Seal leakage for a properly installed and maintained mechanical seal is in the range of 1 to 10 cc/hr.

Zero margin for pump wear is justified by the limited pump run time and the quality of the pumped fluid.

Further, Technical Specification:, 3.7.1.2 and 4.0.5 require. frequent opera-tional testing of these pumps in accordance with the ASME XI code.

The testing program and data offer a complete pump history which will accurately predict abnormal wear, and identify maintenance requirements due to wear.

The head margin availatile for recircula tion flow and lube-oil cooler flow is listed for each pump cn Attachment A.

The vendor's suggested flow'to each lube-oil cooler is 25 gpa which comes from the first stage of each pump with the resultant. loss of flow at the discharge being negligible.

Recommended shut off recirculation flow for the turbine pump is 35 gpm, and 20 gpm for the motor driven pumps.

1 As can be seen on Attachment A and from the informa tion above, there is cde-quate flow. to the steam _ generators from the auxiliary feedwa ter pumps for

=

-North Anna Unit 2.

- ~

ATTACFMENT A AUXILIARY FEEDWATER SYSTEM DESIGN BASIS NORTH ANNA UNIT 2 Calculated Head Pump IIead Reg'd for Flow at 340 gpm Pump Mark No.

of 340 gpm (certified curve)

Head Margin 2-FW-P2 (Turbine) 2949 ft 3116 ft @ 4100-rpm 157 f t 2-FW-P3A 2774 ft 2900 ft 126 ft 2-FW-P3B 2845 ft 2900.ft 55 ft i

a N-i 4

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V aW?

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