ML17258A291

Forwards Final Evaluation of SEP Topic IX-3 Re Station Svc & Cooling Water Sys.Evaluation Will Be Basic Input to Integrated Safety Assessment

ML17258A291
Person / Time
Site:	Ginna
Issue date:	11/03/1981
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Maier J ROCHESTER GAS & ELECTRIC CORP.
References
TASK-09-03, TASK-9-3, TASK-RR LSO5-81-059, LSO5-81-59, NUDOCS 8111050435
Download: ML17258A291 (22)

Text

Docket No. 50-244 LS05-81-OE f Mr. John Maier Vice President Electric and Steam Production Rochester Gas and Electric Corporation 89 East Avenue Rochester, New York 14649

SUBJECT:

SEP TOPIC IX-39 STATION SERVICE AND COOLING WATER SYSTEMS, GINNA

REFERENCE:

Letter, J. Haier to D. Crutchfield, Same Subject.

Dated August 21, 1981 Enclosed is a copy of our Final Evaluation of Systematic Evaluation Program Topic IX-3, Station Service and Cooling Water Systems.

This assessment compares your facility as described in Docket No. 50-244 with the criteria currently used by the Regulatory Staff for licensing new facilities.

Your comments on our draft evaluation have been incorporated as. we deemed appropriate.

Our comments regarding your submittal are as follows:

1)

Comments 1, 3, 4, 7, 9 and 10-General These comments were accepted and are reflected in our final eval uation 2)

Comment 5 -Seismic design of CCW makeup su jply.

While we did not agree with your comment we did reword our concern for the purpose of clarification.

3)

Comments 2,

6 and 8 SWS r equirements.

Based on your comments we have re-examined the SWS pumping requirements and have deter-mined that the present SWS technical specifications do not ensure that the specified required SW flow would be available, assuming loss of one diesel generator.

Our evaluation has been modified to reflect this finding.

OFFICEI SURNAME/

DATEf Bfii050435 8iii03 PDR ADOCK 05000244

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o P

PDR NRC FORM 318 (10-80) NRCM 0240 OFFICIAL RECORD COPY USGPO: 1991~960

4

~ 4441%,P 4

1 L<< t l

'414 34 Ige g 4

~

~ 4 I<<

I'

'I I

I ~,

I 44344 li 4 14 g

~

~

~

C' 4 ~

~4 I

<<8444 4

4 I

~\\

g a

~

4

~

~ I'4 Ill u 4

I" I

4

)$4) 41 41

~

4 I

4 li 444 14 ~

4

'<4 ~

4

~

~

~

V I

~

~ I I 44 ~ 114 ~

~ ea

~ t' I ~ 4 ~

a

~ ~

~ I ~

"341 ~I' Ia -I ~

I~ ~ I

-I ~ ee ~ ~I 441 ~ ~ 1.1

~ ~ate I ~ ~ ~

~ ~ I

~ 11.

Pe ~ -14

This evaluation will be a basic input to the integrated safety assessment for your facility.

This topic assessment may be changed in the future if your facility design is changed or if NRC criteria relating to this topic are modified before the integrated assessment is completed.

Sincerely.

Enclosure:

As stated cc w/enclosure:

See next page Dennis H. Crutchfield, Chief Operating Reactors Branch No.

6 Division of Licensing OFFICES SURNAME/

DATE IW SEP SBrown.

10//

/81 SEP:SL BHermann 1 0/ad'/81 SEP:BC 1N'us's'e IQ-'

0/OI /81 DSnas Ger 10(A/81 ORB¹5/C Cr'u't'c'fi7i'el'd'

$/og/81 gA: DL 1 0/gi/81 NRC FORM 318 (10.80) NRCM 0240 OFFICIAL RECORD COPY USGPO: 1981~960'

0 1

~ tt '

t,l

~

.) f. -" l I

Nr. John E. Maier CC Harry H. Voigt, Esquire-,

. Nr. Thomas B. Cochran LeBoeuf, Lamb, Leiby'nd MacRae Natural Resources Defense Council, Inc.

1333 New Hampshire Av'enue, N.'.

'725 I Street, N. M..

Suite 1100

Suite 600 Mashington, D. C.

20036 Mashington, D. C.. 20006 Mr. Michael'lade:

- 'U. S. Environmental Protection Agency.

12 Trailwood Circle. '.

Region II Office...

Rochester, New York 14618,

-..ATTN:

EIS COORDINATOR 26 Federal'laza

.Ezra Bialik

. 'ew York, New York..]9907

~ Assistant Attorney General Environmental Protection BureaU

Herbert Grossman',

Esq.',

Chairman Hew York State Department of Law Atomic Safety and Licensing Board 2 Morld Trade Center U. S. Nuclear Regulatory Comnission New Yor'k, 'New. York- "l0047.==:

~"

'-"- Mashington, -D. C.

-20555 Jeffrey. Cohen Dr. Richard F. Cole Hew York State Energy Office Atomic Safety and Licensing Board

-. Swan. Street'uilding.

U. S. Nuclear Regulatory Comaission Core 1, Second Floor Mashington, D. C.

20555

'mpire State Plaza Albany,. Hew York 12223 Dr. Eveleth A. Luebke-

'tomic Safety and Licensing Board

~

Director, Bureau of Nuclear U. S. Nuclear Regulatory Comnission Operations Washington, D. C.

20555 State of Hew York Energy Office Agency Building 2 Empire State Plaza Albany, Hew.York 12223 Rochester. Public Library 115 South Avenue Rochester, Hew York 14604 Supervisor of the Town of Ontario 107 Ridge Road West

Ontario, Hew York 14519 Resident Inspector R. E.=Ginna Plant c/o U. S.

NRC 1503 Lake Road

Ontario, Hew York 14519

0 i'

4 4

p

~

~

f 4

l r

~

r r

4 J<<

t 4

4 4.

4 ~

~

r

~

~ r r <<r>rr

~

~

= r r

4 r

rt 4

4 r v

SEP, REVIEW

~

OF STATI'ON SERVICE AND COOLING-r

~

WATER SYSTEMS

'-": TOPIC IX-3 FOR THE R.

E.

GINNA NUCLEAR POWER PLANT t

~

r

~ v

~>

~ ~

~

I,

'I

(

IliTRODUCTI Ot<

1

rTHe safety objective of Topic IN-3.is to assure that the cooling water

-'systems have the capability, 'with. adequate margin,.to meet the design

'objectives and; irn particular,to assure that:

\\

a.:- systems are provided with adequate phvsical seoaration'such that there are no adverse interactions among those systems under'any mode of operation

'.=

. su ficient. coolino=water'nventory has been provided. or.

thal adequate provisions for makeup are avail abl e; tank over f1 ow,.cRnpot bb; r'el eased to the environment, without

.monitoring and unless ihe level" of radioactivity is within

=a ceptable limits-

-"..'d=.-::vita1 equipment necessafy f'r.achieving a controlled aod safe shutdown is not flooded due to ihe failure of the main condenser circulating i~'ater system.

V ~

REVIEM CRITERIA The current 'criteria and guidelin'es-used to determine if the plant systems mee. the.opic safety. objectives are'-those provided in Standard Review Plan ',SRP)

Sec ions 9; 2. 1,'tation Ser'vice 5!ater System",

and S.2.2,

~e"c:ol AU>;iliarv Cootino hater Syste..s."

r r

I I

iV

~tr rir rr&

s canoe G I C

I

=ach o7 -Tn I

v

)I t

(

~

I

~ I VVI V 'I,,I V

I' r

I

~

I C,

l 5 C

~

IC I

I I

I C C ~ I ~

4r v

I reM Ted Topic l epo)

Ls coil L,c i..s ne L,o NV

~

I cvo>" oup/1cat)on OT

~

~

• v 5

I I I vv II' "heno=:ena

. II-3.B.l - Floodino of Equi.pment

. III-3.B - Floodino of Equipment (Failure of Underdrain System)

VI-7.D - Flooding of Equipment.

(Lono Term Passive Failures)

. II1-3.C - Inservice Inspection oi!Rater Control Structures III-4.C - Internally Generated Hissiles III-5 bass and Energy Releases (Hioh Eneroy Line Break)

VI-2. D - hass and ineroy Releases I I 1 Sei smi c Oual ification

~ 1

. EV III-12,- Environmental guali! ication,.

VI-'7. C;1 - Independence of'-Qnsite>> Power YII-.3 - Systems.,Required..for.

Sa<e..Shuido>>g I

"VIII-2'.-"'Diesel Generators:

". !:'--.:--':lv:

~

~

~: <<-'-IX-1 -. Fuel'Storage t V

~

r>>

- lN-6

--. Fire 'Protection

= -'--'-'

'"" -'0 i-I r I

The folio>>ino topics'r'e de'p'endent,on'he.

prese'nt topic 'nonfor!I!ction -for

. C.'opp.l et'i on".

..',. 3Q-3!-".:Cogiainr!!en 'ressur>>e,a'nd Heat'Remova'1"Caoabjlity"'.,","""-

~ 'i'>>

~

.~

..>>."I>>

~r l<'I~>,i i",'l: >>I

>>V a n>>l

~ >>

~

>> '"4.w>>"

l II!

~ l.., 'l " I4 i!=I!,o

=

.-4d>>i':

-::-"'"IX-5 -"'3 en+i at) on" Systems

':.=-:-

-'. -.=-;"::-.-;-"-...:= i: !:.:..',.; >>...:.i:,

,"'-3 W.-7 -" ReaCtOr COO1 an.i PjTmp -Rotor Sei'Zure I

r REVI EM GUIDELINES.......

In "additiOn tO the OuidelineS. Of SRP.SeCti!OriS..;2 1

a!nd

!2::2,', in.

<<det~er'-',",ining which sysiens to evaluate under this topic the staff u!Sed the definition of systems

.important

~0 sa ety" proi!ided in Reference" 1.'he.

defi-niti'on states" systems;.important to sa!ety are those 'necessary.

to ensure.(l) the intecrity of the r'cactol coolant pressure boundary*,

i2 j "Ie ccpacbil ity to shu coivn i.

I 1 ecct.or cnd r!=intc in 1 i. in! c safe c".-.='ii-;icn or ~3i the

)ED=-~~-:..!itv to creven-..

Or.-.'Itiica-.e i.h c"..seou n(-s 0,

co 'ent 't COU! C E'Su!

I n!

" t9nt!c I OI S I I.e E):posut es CO:I" a'. c

! e

~0 t'hie guicelines of

!Q CFR Par'00, "Reactol SitE" Criteria."

5-iree i;e con!s !der S<<ntiic!

Svste~>>..s as Svs I ems or por!.pons OI" Sys;ems important.

-o I,I'I loil

>> 0

~

I

~'l

>>I'C!>>>>>> ~

~

I 0 P

(

r, r

~

~

I e>>

v C,

~

>>>> r' (

rp v i>>

~Iv>>

1'Ic a.

>> I

>>>>. >>>>>>'i I

i'0 v'a'v'c '~

I I

>> v'

~ c vL ~

>> c>>

I el'I I Oc>>

\\

I J

c "ditii0 cJ COO I '3g>>'cfear S) Ste

.S ti c cl e 'mgol ta lt, 7G Sc! e!.y. -'-

-'i.-'. l'AI!0!i In the course of this topic review, the s.aff considered the need to evaluate the Air Conditioning Chilled lJcter System and the Reactor l'!akeup h'at8r System.

The Air Conditionino Chilled li'ater System, which supplies water to ventilation,coolers in the Control Room and Service Building, may be an essential system (for Control Room cooling); however, it in'.1 not be evaluated under this topic until the completion of SEP Topic IX-5, "Ventilation Systems" for Ginna at which time the safety sionificance of the system will be ascertained.

The Reactor Mater l"akeup System was

~

~

Reactor Coolant Pressure Boundary is definEd in 10 CFR Part 50 550.2(v).

~ Q I V:1, considered for its function of supplying water to the -reactor coolant pump (PCP) standpipes.

Reactor makeup water is supplied to the RCP standpipes on an intermittent basis, when the standpipe loses volume-due to evaporation.

Its purpose is to supply pressure downstream of the second stage RCP seal and thus iorce som minor amount of water up through the third stage RCP seal.

This function is not considered important to safe'ty.

The 'systems which were reviewed under this topic are -the Component-Cooling Mater System and the Service Mater System.

.The. Spent Fuel.

Pool. Cooling System is, discussed in the SEP review of Topic IK-1;-

"Fuel. Storaoe.".

A

>/4

'Component Coolino Mater System The Component 'Cooling Mater (CCM) system removes heat f'rom various plant syistem>s and components and transfers this heat to the Service Water System.

The heat loads on the system are:

1.

Residual Heat Removal (RHR) heat exchanoers 2;

Em rgency Core C6oling System (ECCS) pumps =-.

a.

Residual Heat Removal pumps b.

High Pressure Saiety Injection pumps c.

Containment Spray pumps ar~cr f'nnl-r~

PIJ

!>On-regenerati V

neat eXCn>an'gel

'I ~

I I

I A>>

.BaCcOI -5U=- 0

7. CoollnG

~aCS

~

~

G

>'!c

~

~

~ >

c xc

~ ic

~

~

, B.

~a>m>pic

heat, cxchainger 9.

Boric Acid evaporator 10.

i~taste gas

'compressors 11.

Waste evaporator Duri>ng normal plant operation, one CCM pump and one CCM heat exchanger are in operation, and chey can accoHUodate the heat removal load on the system.

Bptk> pumps and heat exchaneers

~~e normally used for a plant.

cooldo>>'n; however, lf one pump or one heat exchanger is not operable, safe operation of the plant is no:. a>fected, but the time to cool the plant ls extended (Reference 2).

CCM pump A and B recei>ve electrical power from 42">

V buses 14 and 15, respectively.

The staff reviewed the heat removal requirements of the CCW systems during~

post-accident conditions.

The accidents considered were tlie Loss of

'oolant Accident (LOCA) and the Hain Steam Line Break (MSLB)'nside

'.containment because these events result in the greatest potential accident heat loads on the CCW system.=

The Containment Fan Coolers are also discussed here because they complement the CCW system in-the post-accident containment heat removal function.

Section 14.3;4:of Reference 3 provides an analysis of containment integrity following a LOCA.

Some part of the energy available for release to the containment.='-.

must be removed to prevent exceeding 'the containment design pressure

"'imit.*

Energy is removed by the fan *coolers and the Containment Spray (CS)..system.

The fan coolers transfer heat from the containment '=

atmosphere to the Service Water System.

The CS system removes heai!= =;-,: -.-..

from the containment'y spraying cool water directly into the containment

.atmosphere.

TMs water, now-.heated,'grains to the containment sump.

The heat is then transferred-to the CCW system through the Residual Heat>ss Removal (RHR) -heat exchangers when the containment sump'luid is pumped, by the RHR system, back to the CS system during the post-accident recircu-lation mode of ECCS =operation.

Two fan coolers and one CS pump are"':='upplied power from separate 480 V emergency buses.

The'inimum combina-

~

of-containment cooling systems occurs after a postulated loss of-offsite power and the 'failure -of. one of the two emergency diesel generators:

This minimum combination (1

CS pump and 2 fan coolers) was the combination

=

analyzed in Reference 3.

Using the design parameters of the CS, CCW, and

SWS, shown in Table 1, the containment analysis of Reference 3 concluded.

that the heat load which must be removed from containment can be accommodated by the CS,and fan coolers given the assumed failure of either diesel generator.

During the initial period of energy release to the containment, the heat is absorbed in various passive heat sinks inside containment.

When. the containment heat removal systems are initiated, the SWS is the first cooling water system to assume a heat load since the fan coolers start to remove heat directly from the containment atmosphere.

At the design heat removal rate of the fan coolers (Table 1), the SWS temperature at the cooler exit is 174'F which is. well below the design temperature, limit of 200-'F.=

The.-'eat removed from the containment by the CS system is collected in the containment sump until the recirculation phase of the accident commences.

During recirculation, the sump fluid is recycled via the RHR system (and the RHR heat exchangers) back to the CS pumps for reuse.

• The SEP will reevaluate the post-accident energy balance in. containment under Topic VI-2.D, "Mass and Energy Release for Postulated Pipe Breaks Inside Containment."

For the MSLB inside containment

event, the amount of energy. added to the containment should be no greater than that added for the LOCA case.

(Ongoing SEP reviews will verify that the assumptions used to determine the magnitude of. energy addition to the containment are acceptable.)

Because the safety injection flow to the reactor coolant system'ould

• not. be available as a heat sink inside containment following a MSLB, the containment sump would be filled by condensed superheated steam from the MSLB and CS water

. and a higher sump fluid temperature

would,

.be achieved'arlier.,in. the MSLB case;th'an in the post-LOCA case.. This. :::

would not affect the heat load. on 'the CCW system, however

- because,'f-:

recirculation of, the containment sump fluid were necessary,- it would not be initiated by the operator until much later into the MSLB accident sequence when containment sump..level would" be approximately equivalent

'o the level when recirculatio'n would be-'initiated following a IOCA'.- ',"

Giyen the smaller. heat release..to containment. following a MSLB and.

approximately,.equal sump,lev'els

'at the start of'ecirculation following'.;.-

both the MSLB and LOCA, the heat.load on the-'CCW'system is expected'-.to be no greater than the heat load following a LOCA.

t E

~ J

)

-Cl' C

V

<<I=I

.Kn the post-accident. case, the 'potentially most limiting single;failure

'rom the standpoint of', CCW high temperature would be the failure in the open, position of the motor operated CCW supply. valve to an idly RHR..:.=-

heat=exchanger.

With half of-the'CCW flow diverted to.an idle.heat' exchanger',

the temperature of the CCW at the exit of the active heat exchanger could exceed 200'F.

This condition can be remedied'y the control, room operator by increasing RHR heat exchanger by-pass flow and thus reducing the energy removal rate of the active RRR heat exchan'ger.

The fan coolers have the capacity and would pick up the additional heat load from containment'o post-accident realignment of the CCW system is performed by the operator except for the opening of a CCW supply valve to one RHR heat exchanger at the start of recirculation and closing, or verifying the automatic closure of,'he isolation valves to the service inside containment.

These actions can be performed from the control room.

The

RHR, CS, CCW and SWS pumps and valves are powered from the appro-priate emergency buses such that a failure of one bus would not pre-vent the operation of the systems as analyzed in the post-accident condition.

During.

normal CCW system operation, single active failure could prevent flow to; the services inside containment (Reactor Coolant

Pumps, Excess Letdown Heat Exchanger, and Reactor Support Cooling).

Loss of flow to the containment services requir'es prompt operator action to prevent damage to the Reactor Coolant Pumps (RCP).

Damage to'n RCP from loss of I

coolino flow could result in pump seizure and cause a loss of flow accident.

(The consequences of a postulated RCP seizure are evaluated as an SEP Design Bas,is Fyent-.)

Plant procedures'eguire (Ref-10) the operator to trip the reactor and then the RCPs within two minutes following a loss of CCW flow to the pumps or before RCP motor bearing temperature reaches 200'F.

Plant shutdown following reactor trip is in accordance with established emergency procedures.

Loss of CCW flow to the excess letdown heat exchanger or reactor supports does not require immediate I

operator action, but, if CCW flow cannot be restored to these components, the reactor plant mgst eventually be shut down.

I I'solation of individual leaking components is accomplished, with the

'exception of components inside.,containment, by. manual valves..

Also, although the CCW'pumps-and heat exchangers are redundant, they are connected by single pipe headers whose failure could di'sable the system.

However, at the operating pressure and temperature of. the system (100 psig,'00'F)

(see Ref ll) a passive failur'e 'would most probably resiilt 'inc a 1'eak rate 'which the staff estimates to be 210 gpm using the methods of.~

Reference 5 for a 10" pipe.

The normal volume of water in the surge

'ank (1000 gal.)'ould provide the operators with about 5'mi'nutes at a-l,eak rate of 210 gpm to stop a leak from the system.

It is improbable" that the operator "could act within this time period, and it is possible

. that the leak may',be in an unisolable portion of the syst'm'., If a loss" of.the CCW systems occurs during normal plant operation, the licensee has an operating procedure that directs the. operator to shut/own.the remactor and commence decay heat removal u'sing the steam generators with natural circulation of the reactor coolant system.

If CCW cannot be'eadily.

restored, a plant cooldown would be commenced.

For a cool.down-with.no CCW, the. cooldown method and system described in Reference 6.

(wi4h Ch'e exception of the CCW and RHR systems) would be 'ava'ilable, and the licensee has proposed a method to achieve cold shutdown conditions independent of the CCW and RHR systems using the steam generators (Ref. 9).

Loss bf the CCW system during post-accident operation was considered in the. Provisional Operating License review of Ginna, and it was concluded that the RHR pumps could continue to operate to recirculate containment sump water with decay heat being removed by the containment fan coo) ers.

However, because the CCW system cools the bearings and lubricating oil coolers for the RHR (and other ECCS)

pumps, these pumps would not be avail-able to recirculate the sump water.

Current criteria for piping system passive failures do not require the assumed passive failures of moderate energy systems (3i ke the CCW) under post-accident conditions, alihough system leaks are assume" Pef. 7).

Therefore, the CCW 'system makeup

~

.capability-should be capable to cope wiin normal system leaKage ln post-accident operation.

\\

We also considered the effects of'such a loss of CCW during a cooldown of the plant with the RHR system operating.

In this case, with the reactor vessel head installed, the RCS temperature would rise to greater than 200'F and decay heat could continue to be removed via the steam generator atmosphere relief valves using natural circulation.

Steam generator feed would be accomplished by the Auxiliary Feed System (AFS).

The plant could remain in this condition while CCW repairs were made.

For normal decay heat removal when the reactor vessel head is removed, adequate cooling can be provided by keeping the core flooded.(using various systems such as RMR and CVCS)swhil.e, repairs are made to the CCW piping.

The CCW system is accessible for repairs and can be filled with water in less than two hours after the repairs are completed starting I

with a completely drained system (Reference 3, page 9.3-18).

During normal.and post-accident operation, thermal expansion and contrac-tion of the CCW system liquid is accommodated by the CCW surge tank, and leakage into or out of the system can be detected by'surge tank level, changes.

High and low surge tank levels are alarmed in the control

room, a'nd a radiation monito'r and alarm alerts the control room'operator to the leakage of radioactive fluid in'to the CCW system from components which contain reactor coolant.

The surge tank also maintains a positive suction head on the:

CCW pumps during normal and post accident operation.

Makeup water-to the CCW'system is supplied by either the demineralized water system or the reactor makeup water system via local manual valves 'i'he auxil.iary building.

The makeup rate is sufficient to accommodate-system leakage however,.the, seismic-classification of the.

CCW makeup'supply.-

.'system is not sufficient to assure the availability of makeup water following a seismic event.

The licensee may be required to provide assur-ahce that adequate CCW makeup can be supplied following an earthquake.

P Based on our review of the.

CCW system,'he'safety related functions are to provide cool'ing '=for. the RHR.heat 'exchangers, ECCS pumps,

RCPs, and reactor support cooling pads.

Of these functions only RHR heat exchanger and,ECCS pump cooling are considered to be essential functions; Plant"'-

procedures provide adequate protection from the effects of losing the other safety related -functions.=

V. 2 Service Water S stem The Se'rvice Water System (SWS) circulates water from the screen" house on Lake Ontario to various heat exchangers and systems in the containment, auxiliary and turbine buildings.

These. buildings are Class I structures except for the turbine building.

The system has four pumps, three of which are in operation during normal operating conditions.

As descri bed in the previous CCW section, two SWS pumps are required to remove heat from components under post-accident conditions.-

The SWS piping is-arranged so that there are two fl'ow p'aths to the re-dunda'nt "critical"*loads identified in Table 2.

Another header supplies various "non-critical" loads (see Table 3).

The "non-critical" loads are automatically isolated from the "critical" headers by redun-dant motor operated valves when a reactor safeguards actuation signal

~ occurs.

Redundant motor. operated isolation valves also automatically secure SWS flow to the air conditioning chill water system, circulating wate'r pumps, and screen wash supply on a safeguards actuation signal.

During normal plant operation, the SWS supplies flow to all loads except the standby auxiliary feed systems.

During RHR operation for a normal'lant cooldown, almost all "non-critical" loads may be removed from the SWS, if necessary.

Following a safeguards actuation signal, the"SWS'upplies'"all "critical" loads except the backup feedwater supply to the auxiliary and standby auxiliary feed

systems, which require operator action'to receive SWS flow.

lt V

To overcome single failures in the system each "critical" load has a

redundant counterpart cooled by the other "critical" SWS header.'f necessary, an operator could cross-connect the "crit'ical" headers by means'of ma'nual valves to achieve added system flexibility. - In the'ormal system alignment, no.sing'j e.active or passive failure could resul,t in. the'.loss'f

'SWS flew to r'edundant "critical"= loads -excep4-=-.","'=::

for the reactor vessel caviQ coolers which could both be disabled by a single passive failure.- Since'he'SWS is a mo'derate energy 'syst'm,,,-

a passive pipe failure would probably result in a leak rather than a

complete.pipe rupture.'sing the method "described in 'Reference 5, the estimated'leakage for a SWS header is 585 gpm for a 20" header at.75 psig.

Although this leak may. pose a flooding problem, the supply function of the affected header would not be significantly impaired.**

A leak from

. the;2.5" supply line to the reactor cavity'coolers would result in the.'-

loss of about 25 gpm.

This leak rate would not completely disable the

.coolers which-normally receive about 45 gpm of SWS flow. - : - "-,

Leak detection for the SWS is provided...-by. header pressure switches '.and by sump level alarms. in the buildings which house the SWS (see the SEP review of Topic I'II-5.B).

Isolation of leaking components is accomplished by closing manual valyes, in general;

however, remotely operated valves can secure SWS flow to'non-critical." l.oads, each CCW heat exchanger,-.

and the spent fuel pool cooling system.

Electrical. power for the SWS pumps is provided by 480 V buses 17 and 18.

One SHS pump is started on each diesel generator during post-accident generator load sequencing.

, Short term post accident heat loads can be accommodated by one SWS

'ump.

Long term cooling, based on the licensee's Final Facility Descrip-tio'n and.Safety Analysis Report, requires two SWS pumps.

The additional pump is'to accommodate the component cooling load requirements.

The pres'ent technical specifications (See Section 3.3.4.1) requires that two of the four SWS pumps and one SWS loop be operational for plant operation.

However, if this plant were operating'with the minimum number of required'WS pumps and an accident occurred the possibility exists that. only one SWS pump would be available.

This is based on the assumption that one of

• "Critical" refers to a..heat load that the licensee has designated as safety related.

h ll

• • The effects of flooding from pipe leaks have been reviewed under SEP Topic III-5.B, "Pipe Breaks Outside Containment."

June 24, 1980

in t'.he two emergency diesel. generators fails to start.

This is the normal assumption for evaluating post accident systems performance.

The licensee should address this apparent inconsistency between what is required for long term cooling and what is available.

This could be accomplished by analyses and/or technical specification modifications.'n addition to the above, the present technical specifications should

.;be made.more:explicit to preclude the possibility that, during periods when the"SWS is in the minimum mode of operation, the two operating

-SWS.-pumpsare,not.serviced by the same emergency diesel gener'ator.

~ A review of.Licensee Event Reports. shows no recurring problems with

~

operation or maintenance of the SWS (or'CW system).

B'ased on our review of the

SWS, we cons'ider the components supplied

.by the "crittcal",.supply headers -(Table 2) to be the essential loads-

'o'n the system.

'II.

'CONCLUSION Based on our review of-the service-and cooling water systems for Ginna, we have concluded that the essential systems and functions are:

Component Cooling Water:

RHR heat exchanger cooling and ECCS pump cool-ing.

Service Mater System";

All components supplied, by the "critical" supply headers (Table 2)

'e

.have determined that the design of the. above systems is in conformance with current regulatory guidelines and with General Design Criterion (GDC) 44 regarding capability and redundancy of the essential functions of the

systems, except for t'e apparent SWS technical specification inadequacy.

The systems also meet the requirements of GDC 45 and 46 regarding system

'-design to permit periodic inspection and testing.

With respect to the seismic design of the CCW makeup supply systems the licensee may be required to provide assurance that CCW makeup can be supplied 'following an earthquake.

This will be determined in the'nte-grated safety assessment for the facility.

As was indicated in Section III of this evaluation, the spent fuel pool

.cooling'ystem is being reviewed in SEP Topic IX-1.

That review will also address the proposed modifications to the spent fuel pool cooling system.

If the findings of Topic IX-1 necessitate any additional review of the

SWS, it Hill be addressed in the integrated safety assessment for the facility.

0

..- 10.-'

TA""LE 1.

SYST"=t>

DESIGN< PARNlETERS L

Svsl.e /Reierence Parameters Conta.-n ent Spray

':. (Ref=.;-3.Section E;:4');-'~::."",; =-

2-pumps - 1250 cpm each 2

RHR 'heat'xchangers

- 24. HE6 -BTU/hr-.

each (with 1525 gpm RHR 8 160F'and 2780 opm CCM I 100 F)

Componer ~-.Coo.1'ing (Re i. 3, Section

9. 3) 2 pumps - 2980 gpm each 2

CCM heat exchanoers

- 24. 15E6 BTU/hr each (with 2950 opm CCM I 117'nd 5055 opml S'n'S 9 80,.)

"". 'v l cel, w

l 6 q C

~v'>s

~. v) l

~ laC

~

~ l I

ll g,

l mX I il Co,ol er,s (RH..'3, Secti on 5. 3,

6. 3, and. 9. 6)

(at 50'i 0

'2 6

and 4248'pm S4'S)

"Evaluated in Reference 4.

~I 0 ~

~ I IN It'll

11

, TABLE 2. "CRITICAL" SWS LOADS 2.

3:

~ M p Safety injection pump bearings ECCS and cha'rging pump.ventral.ation coolers Component Cooling Water. heat.exchangers 4..Containment.

Fan. Cool ers. (and motor s)

~ '

5.

Diesel genera'tors 6.

Auxiliary Feed System (feedwater supply and pump cooling) 7.

Standby Fuel Pool Cooling 4

--12=

~ ~

TABLE 3.

"ilGN-Cr". 7 ICAL" Si'(S LOADS

~ ~

'3.

e coolers'Plant air co-pressors

~ P

~

t!

I x

5

- S,'

Conden ate and.,heater dry'in:.pdmps----: -"--- ""- "'"=-

~e e

~

I t

~-~

I I

r

-. 'Relay room ai'r cond i~ionin'g-.-"

i e

j I

~ -

1 e

I" Turbine ex itt re- ':..:-:;,,:.': '=,,'

e I

C t ~

V Bus dUC e.

6.

Seal oil Unit I ~

tl r'

~

~ !)r! !'Alt

! CJIL

% ~

sys -"-.,

10.

Turbine oil coolers Feed pvmp oil coolers e

12.

Containnent. pressvre

'estino air cool'er '" "'

I I '

a Cvuiil PliiilPS

~g 14.

Fire booster pump supply 15.

16.

Circulating Water pumps Screen.wash, supply..

17.

Air conditioning chill water coolers 18.

Reactor.Yessel Cavity Coolers=-

19. 'ontainment Penetration Cooling

~ \\

I

~ ~

Sprey Ccnuinment CS To Sl Pumps Q;t,ei CCW Func ions Containmenl Ait Coolers r

RHR HX CCW

~A RHR Cnner SWS Funciions

~ (

CCWHX SWS r4 r

e r,,"> ~

~

~

3.0 RE 2.

FERENCES Regulatory Guide 1.105, "Instrument Setpoints;"

II Fire Protection Evaluation, Robert E.

Ginna Nuclear Power Plant, Unit', transmitted by RG8E letter dated March 24, 1977 (Section 3.2.6).

a 3.

4.

Robert-Emmett Ginna Nuclear Power. Plant; Unit No.

1 Final Facility.

Description a'nd Safety Analysis Report.

Addendum..to the Safety Evaluation by the Division of-Reactor.-

Licensing, USAEC, for the R.

E. Ginna Nuclear Power

Plant, dated September 19, 1969.-.

.; 5.

6.

7.

Bran'ch Technical Position MEB 301, appended to Standard'Review Plans 3.6.2.

SEP Review of Safe Shutdown Systems for the R.

E. Ginna Nuclear Power Plant (SEP Topics VII-3, V-10.B, V-ll.A, V-ll.B., X),

Revision 2, 'May 13,'1981.

Staff Discussion of Twelve Additional Technical Issues Raised by-Responses'to November 3, 1976 Memorandum from Director, HRR to NRR Staff, NUREG-0153, Issue.tl7; December 1976'.

8...SEP Review of Residual Heat Removal System Heat Exchanger Tube

Failures, Topic Y-10.A, dated February 2, 1979.

9.

RG&E letter L. White to D. Ziemann, dated December 28, 1979; transmitting Fire Protection-Shutdown

Analysis, R.

E.

Ginna Nuclear Power Plant.

10.

RG&E letter J.

E.

Maier to D;

M. Crutchfield, NRC, dated January 23,

1981, concerning SEP topics III-l, VII-2, VII-3, Enclosure 2.

HRC letter D. Crutchfield to Leon 'D. White, Jr.,

RGE, dated June 24, 1980 transmitting review of SEP Topic III-5.B, "Pipe Break Outside Containment.'"