ML19340C783

Forwards Revised Response to FSAR Question 211.86 Re Residual Heat Removal Sys Requirements,In Response to NRC Request

ML19340C783
Person / Time
Site:	Summer
Issue date:	12/11/1980
From:	Nichols T SOUTH CAROLINA ELECTRIC & GAS CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
NUDOCS 8012170431
Download: ML19340C783 (31)

Text

/

l SdVTH UAROt.INA bbLr?!c 8 gas COMPANY pcST O F F4C E son 764 COLUMBIA, SOUTH CARO LIN A 29218 T. C. NicHots, J n.

v,ctr.t,,c....o.o.cn n,*

dccembe3 al, 1980 3

Nocttse 0.tm.ticos Q

o IS a

L'.;j v> G

~,.

c..o m

Mr. Harold R. Jeaton, Director d

.Uy 4

Office of Nuciecr Reactor Regulation M

y

~

U. S. Nuclear Regulatory Commission y

M Washington, D.

C. 20555

Subject:

Virgil C. Summer Nuclear Station Docket No. S0/395 Reactor Systems Branch - FSAR Open Items

Dear Mr. Denton:

As requested by Mrs. Amita Gill of the reactor systems branch, South Carolina Electric and Gas Company, acting for itself and agent for South Carolina Public Service Authority, provides forty-five (45) copiec of a revision to the response to FSAR question 211.86.

This revision, as dis-cussed with Mrs'. Gill, should provide-you with sufficient information to resolve the open issue regarding RSBS l.

If you require additional.information, please let ua know.

j Very truly yours, a

ff

/

I T, C. Nichols, Jr.

RBC:TCN:rh

Enclosures:

cc:

V. C. Summer w/o enclosures G. H. Fischer w/o enclosures T. C. Nichols, Jr. w/o enclosures i

E. H. Crews, Jr.

g O. W. Dixon, Jr.

j D

A. Nauman O. S. Bradham 4

W. A. Williams, Jr.

A. A. Smith A. R. Koon

~,

R. B. Clary J. B. Knotts, Jr.

J. L. Skolds B. A. Burscy j

M. A. Barnisin j

R. Faas NPCF/Whitaker 4,

File 4

4 -3 [

RO19374

Ed 211. CG

[';

bur re sponse to quest ).on 211.12 is unacceptabic.

The USNRC Regulatoiy Requ hements Review Committee has recently approved a new staff position (BTP RSB 51) for the residon1 heat reruo vn i system. The technical tequire-cents of this nosition for your pisnt ace ciescribed below. Wut re sp ons r.

..a these..reqt i tements shoulci ha in

~

sufficient detail to enable the staff to review your com-pliance.

1.

Provide satety grade steam generator dump valves, operators, air and power supplies which meet the single failure criterion.

l 2.

Provide the capability to cooldown to cold shutdown in approximately 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> assuming the most limiting ringle failure and 1.oss of off site power or show that manual actions inside or outside containment or return to hot standby until the manual actions or maintenance O

can be performed to correct v

the.failpt a.

acceptable alternative.

_ providcu a,.

3.

Provide the capability to depressurize the reactm coolant system with only safety grade systems assuming a single failure and loss of offsite power or shov N

that manual actions inside or outside containment or remaining at hot standby until manual retions or repairs are complete provides an eceeptable alterna -

A 3

tive.

4, Provide the capability for borating with only

~

safety grade systems assuming a single failure and

~

loss of offsite power or show

]

that manual actions inside or outside containment or remaining at hot standby until manual ac tions or repairs are completed provides an acceptable alternative.

N V

211.86-1

-?E

-m M NDMENT 0

'hf SEPTEMBER, 197

~

g**g *9"%

ft f

.f.

. j \\.

W JU cj

.a 5.

i'rovide the syc:eur and component design teatures necessary for the prototype testing of both the mixing of the added borated water and ihe cooldown under natural circulation r.onditions with and without c single.fai. lure of a steam generator atuospheric dump val ve:.

Toese.teste and analyses uill he used to obtain informatiisa oc cooldown times and the con:e-sponding AEW requirements.

6.

Cer.:mi t to providing specific procedures for cooling 3

down usicg natural circulation and submit a summary of these procedures-,

7.

Provide a seismic Category I AFW supply for at lear.t 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at Hot Shutdown plus cooldown to the DHR sysLem cut-in based on the lougest. time (for only onsite or offsite power and-assuming the worst single failute),

or show that an adequate alternative seismic Category I source will.be available.

m 8.

Provide for. collection and containment.of RHR pressure relief or show that adequate alternative methods of e

l}

disposing of discharge are available.

N

RESPONSE

Ref ace w't%

h seer A l

~

War a a d r'.sient eve s,

en th r cto tri, it i impo t t oh e r iable mec nis for 'raasf rin dec.

heat rom e

7; v

r cto t th envir men whi at nea norm op ratin teca a-ur It s des

• abl to n e the abili y to ems' at hot andb n tio for a ind rinite perio of t'me ae it r duces the ssil lit' c

cor melt nd p ovides great r fle.ibili y in deali g wi+ i ot r

v over tinc

.ndit'ons.,

I

.- g j

"4 211.86--2

AMENDMENT 12 J YRBRUARY, 19/9

_-e Tfw ec[G coydowpn e

be condi ions whi wo td qu' e ev nto goi g f.

.o er

.o it ~ong t c olin-wi t

RH sys om o c d utdown fog inspe ion and-epatTs.

In eepi g wi i th abo e a

/

/

/

yn espoote tojQuesti a n.86 nd R$8 5_b_ ' ici fo lowi' e i nrov dc

.o 1.

Pr4 vide sa fety gradc oteam generatc,t.demp. valves, operai ors, a n and.

'}

power 'supplica khi@ meet :the. singic failtn'e criterion.

1 One safety grade-steam-genetcLor : power operated relief velve c.

'5 provided for each of -the thre.c-steam generators.

Safety grade remote operators and power supplits. are not required since hot standby can be achieved and maintained using the safety grade steam generator safety valves. The steam generator power operated relief valves are provided with handwheels and can be operated locally to j

permit plant cooldown..See tlie cold shutdown scenario and single failur.e evaluatiou prsncided b_elow.(Part II, Removal of Residual Heat).

2.

Provide.the capability to cooldown.t.o cold shutdown in approximately C.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. assuming the -cost. Licdting single failure and Inss or off-site power or show that manual:aetiens inside or outside containment or return to hot standby until the manual actions or maintenanc.e can

.be performed to cor rect the failure provides en acceptabic alterna-tive.

%e If a condition occurred requiring cold shutdown of.the plant de. sign features permit the msintenance of a hot standby condition for an indefinite period of time. Tne plant is capable of being cooled via I

natural convection and reaching RHRS initiating conditions including the time required to perform any manual actions.

[

3.

Provide the capability.-t'o d' ssurize the reactor coolant system with only safety-grade syst.jms assuming a single failure and loss of offsite power or show that manual actions inside or outside contain-cent or remaining at hot standby until manual actions or repairs are

- )

complete provides an acceptable alternative.

c3 c3 c3 f I f.

I 0

0 0 ALhitL 2

q

.o A a

l\\

W 211,86--:i AMENDMENT 12 S

b

.-__-_._---_-.~-...,,.~.,.n,

~

J hk

'1he plent can be maintainco in a safe hot'standoy condition while any requi;.cd manual actione tre iaken, 3ee the c.old shutdown

'9 scenario and. single failure evaluation provided oclow (Part IV, Depre.s uu s ization).

r m-4.

Providt; the. capitbility tor hcu at. tog with-on)y calcty -grade systee<4 assuming.a single failute sud loco caf offsite poncr or abow that

.manpal' a ct tons. inside ' colaide cantainment ot: remaining at hot standby uutil wauual artions oc tepairs are completed provides cc.

]

acceptable alternative.

The plant-can be maintained in s aafe he,L:st audby.. condition while any tequited manual actions are taken. See the cold shutdown scenario:and single' failure evaluation'prov-ided below (Part III, Boration and Makeup),

5.

Provide the system and component denign features necessery fer the prototype testing of both the mixing of the added borated water and the cooldcun under. natural ctreulation conditions'with and without a single failure of a steam T,enetwtbt atmospheric dd.:p valve. These

~

tes tc and. analyses, will-be. used-tog obta in information ca cooldown times and the-ccetespondinh ER rr.quireuents.

l 84 The Virgil C. S.u=mer. Nuclear Station. design provides the capability Q

~

for conducting natural circulation oneiheewn tests. The ability of at kor suhy co"br'o"*

natural circulation to remove decay heatAhas been. demonstrated at L "- H ic-parr of fhe few p e u.a c.-

-tese

' 14 m 3.

L '. I.

'Jc ;p ro p ea n.S Tests have s t u ted f

P a~ti asa u.

shown natural circulation flow to be more than adequate to remove decay heat. [0tb.r F st ng! us ci it ieu re sur ze wa er ca or. l I

s ne ul d e dt ici ar te s

io t

tF s t' rt o

th.

ar j

r il r.er uc car St tio in e er.i a be ent al p

~

s '.2 i r y d ig f a

k' s t

• ngh us p ss tiz w ter r

en to s,

.ie s te ts il be re rer2nt ti'e o co dit on fo d i

M.M th vi cil C.

ciu...er ou ea St ti n.

Rylu.e w uth Ioseer 6 j

i S

'211.86-4'

. AMEND'ENT 14 nmE, 19/9 g

F

}

\\t

... 2 ::._ m :.

-% -- ~--.--

..c.

c-w-

ht re ie o tes

-r ult.

om ' ic s ou a.W sti h

d ou.e u its, de er-in. io wi

1. : b ma'c.. to wh* h r sul,, as ap ica' le t th.

'tr il S mme th.le 5 all i.

to edur s t dev lop ddi ma i

.o

. ti o co do n L mes an cor, esp ndin,em ge y fe 2dw. er equ.rer.nte vil/ be dev

+

16 d.

6. - Commit. to ptoviding specih.e ptocedurcu for ersoling devo usint natural circulhtion.and subinit a

• aummat y of 'these procedures.

Specific procedurea in cooling dom using odtural ci hulation w*11 4

be prepared prior to t he e tactup of the V'.rgil C Suroner Nucleac Station.

A.su= mary of the procedures is provided in the cold stwt-i down scenario.and singic. forlure evaluat'i6a provided below.

7.

Provide' a seismic Category -I AFW supply for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at Hot Shutdown plus cooldown to the RHR system cut-In based on the longest time-(for only onsite er offsite power and assuming'the worst single failure), or show that an adequate alternate seismic Category I

['~'.

source wil-1 be available.

C Sufficient emergency feedwater is provided in the Seismic Category I condensate storage tank.-to permit fouw boots.opetation at. hot standby plus cooldown to RhR itiitition cdnditions.

In addition, a 2

long terrn source of Emergency Feedwater is provided by a connection to the Seismic Category I Service Water System.

See the cold shut-4 down scenario and single failure evaluation provided below (Part II, Removal of Residual Heat).

8.

Provide for collection and containment of RHR pressure relief or show.that adequate alternative methods of disposing of discharge are available.

- O' The RHR relief valves discharge to the pressurizer relief tank, located inside containment.

O' J

G f.

211,86-5 AMENDMENT 12 pl FEBRUARY,.1979 r

U l

b m. _ _.m g g,

,~

f.

y Dh

@d D IV 4,an f

b uI M

bL klrL 2 D '4 COLD SHUTDOWN SCENARIO 4g 8 M,;2t

~ ~ ~ - - -

3 '**.+ 7 The safe shutdown design basis. of Virgil C, Summer Nucicar Station in *

1. P D I/,

• -t.

ty; hot standby.: 'Ihe plant can be caintained. in a safe hot scandby condi-2'S'P a

~TjI tion while manual acticas are taken_ t o permit achievement ot cold 4 *e 2 o'1 j

~1 E g c, g 4

shutdown s onditions following a safe-shutdown earthquake: with loss ut.

], { A-a

%, f c.ffsite power.

Under such tonditions', the plantais capable c f achiev'tng dI2E o ;;,.

RHR. initiation conditions (approxt:nately 3500F, 4?S psia), iucludink 0*E1 C} a t r

=1 1 * *..

the tirac required foc any raanual actions

.To achieve and raaintain cold f s* er a a

g

, shutdown, f our key f unc tions must be perforraed.

These are:

(1) circo-

+$

• k ( ?

1ation of the reactor coolant, (2) removal of residual heat, (3) bot'a-(#

<F c

2 8 LA

(.,

tion and mah.:up, aad (4) depressurization.

• 3 *,7 et O

'SN,

@ r* e d e

,2 E. g, u

~1 [ ' 7 2

E3 I.

Circulation of Reactor Coolant 7

,u s-w. -

g t

e 6 o."

t.o<4 a

Circulation of the reactor coolas b u.tvo stages xu co'oldown frcro hot

.t 4.-

~t ; r c-

+

- a

[7%aNw standby tn cold shutdown.

The first stage is from hot standby to

. s-f, &

en a

S c i* ft 350 F.

During th'.s stage, circulation of the reactor coolant is pro-g.

AE a 2. % C M*+*D vided by uatural cit'culation yttb the rasctor core as the bent source O

sc t

~

t% t%

_e o e

'A g *,

and steam generators as'the hust sink,

.ittom m teace trom the steam e

r.

, (v4 3

• ~

I k *,,, h generators is initialty via che utcam generator safety valvc<: and occirs 2, e,.

1.7 1s2 autenatically as a result of tutbine and reactor trip.

Steam release j*

a t

.c a

, - e,.

c4y foc cooldown is via the steam generatot power operated relie f valvet-[

{yi

$ *. 7 (.

CJ c

.J...

_. The steam generator

'* q 2' }

"J t;

tasjty

[{ h Tg. gpower operated relief valves areAaccessible for-local operationg. The,

44 g%

i a

,g go status of each steam generator can be monitored using Class lE instru-p=M a4

,t 7

mentation located in the Control Room'.

Separate indication channels for

+3

~

+;4

• y f.*

')hhjbothsteamgeneratorpressureandwater.levelareavailable.

N *

. c n t-n.D m

A 'a 8

3 a

g k 'S f Feedwater to the steam generators is provided from the E5ergency Feed-

$o )*

[

water System which hes a 150,000 gallon Seismic Category I condensate E

I vp

'} E.1 7 storage tank as the primary source and two separate scismic Category T 2

o a a e e

a q u

3,. -+ 2 y-piping sub-systems. The first sub-system is composed of tuo raotor-

?

.j 7 e

M.;

driven puaps each powered from a dif ferent emergency power train, and g$

o

• A x

E the secend sub-system incorporates a turbine driven pump which can

+

.v o

?,,.

?

1 1

21L 86-6

..AMF.NDMENT I2 F8BRUARV, i979 i,-

e

.c

.m.w

..go..

~

'recei<e motive steam from either of two steam getierators. There ate additional sources o.t feedwater backup which can be manually accessed.

Initial backup is ptovided by the demineralized. water storage tank, ne

. the G1tered water ritorage tank. Additional backup is from the Seismic Category' l Servicer Rater System. The opetation of the emergency feed watet ' system-ean be ennitored using Class. lE instrumenta'tirm located in

,. che Cont e r 1 roow-The socond. stage mi Reactor Coolant CicFulstion rs from'3300F to cold c

g m

shu tdown. Dia ing 8.his s tage, cicculation of tlie item f or coolant is provided by.the Residual IIcat Removal Pumps.

II. Removal of Residual IIcat Removel of cesidual heat stso has two stages in e cooldown from hot standby to <:ald shutdown..ihe ti, ct stage is (tom hot standby to 4500F.

During t,his stage, the staam generators act ar the raeans of heat removal a

from t he Reac to c.CoM ank -:iyeN.n, st.u t tal ly,. 's te n is i clecor d front ite steam ge.neratots :via the ste.am gcuerator safety valves to. maintain hot standby conditions.

Whe.i the-opetators are ready to begin the cooldown, 7,

the steam generator -power fipcrated reliet valvea are clightly opened by l'ocal operation'vith their.handwheels.

'As the cooldown proceeds, the p

operators will occasionally adjust these valves to increase the amount they are open. 'this allows a reasonable cooldown rate to be main-tained. Feedwater makeup to the eteam generators is provided from the Emergency Feedwater System. The Emergency Feedwater System has the ability to remove decav heat by providing feedwater to all three steans generators for extende~d periods of operation, The second stage is trom 3500F to cold shutdown < During thic stage,

~

s the Residual IIcat Removal (RIIR) System 'is brought 'into operation. Thc:

Residual I! cat Removal liest Exchangers in the RiiR system act as the means of heat removal from the Reactor Coolant System.

In the RIIR IIcat

(:J la J

211.86-7

. AMENDMENT 12 FEBRUARY, 1979 f..--....-..-....-..

J h~

[=

' Exchange r, the residual heat;-it> transierred to the Component Coolt

' (!. x S$/steerwhich transfers the heat to the Service Watcc System. The Component Cooling and the Service Water systems are both designed to Seisciic Category 1.

The RHR system includes two Residual Heat Removal Fumps, and two Residual iteat Re'moval Heat Exchangers.

Each RHR Pump in

.povered trora.di t t'etent cactgency power trains ~ and each RHR Heat

. hchenger. is ccoj ed by a difiennt-Component Cooling loop.

af any

.omponent: 2n one;RhR loop-becomes inoperabic, cooldown of the plant is

'.not:::ompromised; howevei, t he-time for ' coldown would. be cytended.

e 4

)

fac operation of the RRR system can be 6:enitored using Class lE instru-mentation its the Control Room.

IU.

Boration and Makeup Boration is accomplished using portionn'of the Chemical and Volume Centrol System (CVCS).

Boric acid 4 wt. % from the Boric' Acid Tanks is

- supplied' to the suction of the Centrifugal Charging Pumps by the Boric Acid Transfer Pumps.

The Centritugal Charging Pumps inject the borated water into the Reactor Coolant S tora via t he. betwa) charging and/or reactor coolant pump seal injection flow paths. '. The two Boric Acid Tanks, two: Boric Acid "Iransfer Pumps, Centrifugal Charging Pamps and the associated piping sre of Seismic Category I design. There is sufficient boric acid capacity to provide for a sold shutdown with the most reae j

tive rod withdrawn.

The Boric. Acid Transfer Pumps are each powered from di f ferent emergency power trains. The Boric Acid Tank level can be monitored using Class 1E instru=cntation in the Control Room to verify the operability of the boration portion.of the CVCS.

An' alternative boration source is the 12 wt % boric acid contained in the boron injection tank located.in the Safety Injection. System.

This source can be used to sup'plement the boric acid-tank to accomplish bora-4

)

tion, depending;on initial plant conditions.. The contents of the boron n

injection tank can be delivered to the-RCS by aligning the discharge of the Centrifugal Charging Pumps to this tank while the suction is aligned

-[ A to the boric acid tanks.

1 y 2

s

'V 211.86-8 AMENDMENT 12 FEBRUARY, 1979 y-

. - r..r -- -

1 Makeup, in excess of that required for-borationi can be provided from 1

the Refueling Water Storage Tank (RWST) using Centrifugal Charging Pumps

,]

3 and the same: injection fldw paths as described for boration. Two motor

_ operated valves, each powered from different emergency power trains and connected in paralteJ; Sill trEnsfer the. suction;of the charging pumps

.to the.RWSTe Makeup from the RWST can be monitored using Class 1E

,)

instrumentation in the Couttol Room.

W.

Depressurization i'

• * ' I' "' f *

• U y as s e.onf t s h t 0 U !* **t, "

• t'" Y

is

• o o r >< s t l D<pr$surhs5 e -.* n < uor~st stesy tints fe the f r e ss

• n %

• t-e y

i l DepressurizatioE* M h c6mplished using portions of the Chemical and Volume Control System (CVCS).;.Either'4 st, % boric acid :or refueling water can be used as desired for depressurization with the flow path being from the Centrifugal Charging Pumps to the auxiliary spray valve to the. Pressurizer. The Centrifugal Charging Pumps of the CVCS are of Seismic Category I and are powered from different emergency power trains. The pumps can be operated from and their operating status moni-tored in the Control Room. The depressurit.ation of the Reactor Coolant System can be. monitored using Class 1E instrumentation in the Control

.q Room. Available to,the operator are pressurizet 'teve.1 aud Reactor Cool-ant System pressure and temperature.

4.T**strT C

Mn err tiv m ho r' of d prese riza Aon, afte bort ion. ad RC cool I J nt 45 F,

one sts f di har 'ng R S inv nto fro the ressur-h This operat'

.ze to ont inn it v a the pres rize safe y'v/ives.

a 4 ald e ~ te ate with te c,oldow fun tion nd is n'c.plishe by e

2ti izi g t Ce trifun 1 Ch rgin Pum to 'ncreas Re. tor Co lant 3 ste. pr_ssu e to t e sa'ety v ve etpo*

t.

Ae the,afety alves CI re ev inv ntory o co.tainm it, he w.

er in he p essur ier is r pl ced y coo r wa er fr m t Reac or Coo ant ystem oop p' ing

(

co ling the p ssur zer t 45 F.

R actor ool, t Syst m prer,ure nd p

t mpe ature ind P essur*zer evel e n be e ni red us'ng Cla.s 1E

_j

[

~

[n e n-n e tion in the Con ol Room.

1 211.86-9

- AMENDFENT 12

lt i

,.o l A\\

o f

" J]"

' UA\\ #

D 9 70

• D

' T M li\\

1 J

Maintaining RGS TemJerature and,f ressdre Without Letdown P

}

ll

& W 5A!SetY

.'D l..

In performing the:cooldown, thc' operator will Integrate the functionn of i.

heat. removal, beration and makeup, and depressur.1zation so that these

'j v

~

j

- functions can be. accomplish'ed without letdown from the Reactor Coolant.

e

[

Wyatem. 'tsoration,.cooldown, and depressdrization will be. accomplished

,a

{. ' ' '. -

in n'su-it:c 9 short steps arrangcd tn Lehp Reactbr Coolant hystem tetr-1

. ymtur e and ; pres.sure-tad PressurTrer 1cve-1.iu the..decited rel.ation-j

. chipr.- %e *odration i.tiquirement.wlU be evaluated by Lt;c ' opes. ator p>. ior

~

tes initiating.cooldown.ciidcdepressdrization. Based on initial plant conditions, -t he operator may elect to borate using the contents of the 5

boron inject' ion tank sad /c,r -thel contcai:s of the : boric adid tanks. Oce

-ij-the plant is l cooled.to 3500F and -depressurized tJ 425 psia, Residual 3

Heat Removal System operation is i'nitiated and the Reactor Coolant System is. taken to cold-sliutdown conditions.

Tcr demonstrate that boration _and depressurization can be done without letdown, a simpler. scenario can.he used. First.the operators integrate the cooldown andebocation-functions takingfadvantage of the m e.

i - -

2. : ' 2 :M : i 2 S c, g a ; r : =

d' JcRCS'.iuvenan y, cc,ntrac td or> cenul;ing from -the :cooldowns. Finally,.the. operators.use auxiliary spray from the CVCS -to ;depressurize the pidnt to RHRS initiating conditions. The 1

calculation to denonstrate this capability assumes ' worst: case boraHon 4

- requirements based on cot'e,end '.of life /,pe-ak xenon. conditions and the h

following RCS initial conditions following plant trip:

1j RCS Temperature 5570F

)

-S RCS Pressure 2250 psia 5

Pressurizer Water Volume 350 ftt Pressurizer Steam Volume 1050 ft3

.i The cooldown froI:1 5570F to 3500F decreases.the volume of water in

?

s

-the RCS.by'approximately 1250 cubic feete This assumes that the pres-surizer is not. cooled and the water level is maintained at the initial

[,

J' conditionc Makeup -for contraction is supplied by 4' wt. % boric acid a

.w.f i

)

3 211.86-10

.. AMENDMENT 12

. 9't:3RUARY, 1979 I

L

s.

A boric 'cid tank volume of stored'in the: boric acid tanks at 7001..

a approximtely 1130 cubic feet will expand to' approximately 1250 cubic

- i feet as it is heated to the RCS temperature of 3500F. 'A boric acid tauk, volume of'approximately 1000' cubic feet is required to maintain the _

seactor within the technical specification. shutdown requirements at

~

The volume requir'd.for hnration requirements at 3500F is

1500F.

e

• -tess-i.han.the contraction' volume.,a'vailable at 3500F.

'ar calculatet iE.deprecouriention can be acto oliched without letdown and e

.n

) -

_without taking the p1' ant. water solid, it wa assuried that the Pressur-1 iter war at caturated condi.tions with 350 cubic feet of water, 1050 cubic feet 6f' steam, road. the Press'1rizer metal, all at 6530F (2250 f

psia)

It was further assumed that no additional water would be removed from the pressurizer by the cooldown cdntraction. sWith these assump-tions, and including the e-ffect of heat' input from: the pressuris.er

.;.. metal,- it. was ' determined that spraying iu~ ap'proxitiatelp 500 cubic feet of 700F water would provide saturated conditions at 425 psia (4500F) vith a water'. volume of 960.. cubic feet and.a steam volume. of 440 cubic 1

feet.

I Once depressiirizsd N 425 psia, RHRS operation is. initiated and cooldorm h.

is continued to cold shutdown conditions. The co'oldown. from 3500F to

-200 V further' decreases tl e volume of Water' in the RCS by approxi-mately 350 cubic feet. This assumes that the pressurize'r is not cooled

~

f.

and the pressurizer water level is maintained at the level resulting l

from depressurization. Maheup for contraction is. again supplied by 4 wt

% boric-beid. 'A boric acid tank. volume of.,approximately 340 cubic feet a

will expand.to approximately 350 cubic feet as it is heated to the RCS i.

(

temperaturetof 2000F. An additional boric, acid tank volume of only approximate 1p:250' cubic feet is required to maintain the ' reactor within

}

the technical. specification shutdown requirements of 2009F. The addi-

}

tional volume r'equired for-boration'. requirements at 2000Fris less than the additional cotitraction volume available' at 2000F, thus ensuring that tcchhical specification requirements for cold shutdown conditions are satisfied.

[p**g M.L: ~ 3~ f g fGN 2 eo e 0 211.86-1J /c. AMENDMENT 12 [ .: FEBRUARY, 1979 y

4 T Aeld rner E IThe - s ts ft c cul io der rib ab ve de.onst t e' at b aT 'o an epr su zat' ne be ace plir ed w' hou etd en an wit 3 .. c.; 4 o t,cin - ul cre t f th ava' ab] vol .e er ted y th coo own on.[ f r co rac u O Instrumentation Class -1E instfumentation in available in the Control Room to monitor the key s f unctiotic.acsocihted w* th achieving cold shutdown. This instrumen- ) tation-is. discussed in Section 7.5 (Safety Related Display Instrumenta-tion) and includes the following: 1. RCS. wide range temperature 2-RCS wide range pressure 1. Pressurizer water level 4. Steam. generator water. level (per steam generator) 5. Steam line pressure (per steam line) 6s RWST le,rq 7. Boric a' tank level (per boric acid tank) t. Henctor-building pressure 3 -This instrumentation is.suf ficient.to monitor the key functions asso-diated with cold ' shutdown arid to maintain the RCS within the desired pressure, temperature and inventory relationships.-Operation of the N auxiliary systems that service the RCS can be monitored bp the Control Room operatcr, if desired, vi2 remote communication with an operator in the plant. i SINGLE FAILURE EVALUATION l l t I. Circulation.of the Reactor Coolant e l. From Hot ~ Standby to 3500F (Refer to FSAR Figures 5,1-i; 10.3-1, and 10.3 t4) - Three reactor coolant loops and steam generators are provo. lee J 'd, any two of Wich can provide suf ficient natural r-circulation o, q- .o m M e lh o i 3 711.86-12 .A}fENDMENT 1% g rew ~> x r im

E L. tiew to pcovide adequate core cooling. Even with the mose iimiting . :7 cingle f ailure (of a steara generator power operated relief valve), two of the' reactor coolant-loops and steam generators remain avail-able. ( 2. From 3500F tc. cold shptdown. (re.1er.I a F9AR. Figure 5.S--4) Two R11R pumps are provided,.cither.one ofcwhich can provide adequate circu-lation of the teactor co ol ant., i II. Removal of_ Residual,, Heat 1-From Kor. Standby co 'tS00F (Rcfer to FLAR Figuier, 10.3-1, 10.3-4, _ 10.4-16 and 9.2-1, and 3) Steam generator power opecated relief valves - Three are pro-a. vided (one per steam generator); any two of which are sufficient for residual heat removal. In the event of a single failure, .two p6wer operated relief valves remain available b-F.mergency Feedwater Purps 'Cwo. motor di neu and cpe neam dcnen emergency feedwater pumps 'are. prn'vided. _ In the event at ~ a single f ailure, two pumps remain available, either of whict; can provide culficient'feedwater tiow. K c. Flow control valves - Air operated, fail open valves are pro-vided. In the event of a single f ailure of one flow control 5 valve (which effects flow to one steam generator from either a motor driven pump or the steam driven pump) emer'gency feed flow can still be provided to all three steam generators from the other pumps. ^ ' d. Backup source - A backup source of emergency fe'edwater can be provided to the suction of the emergency.feedwater pumps from cither train of the Seismic Category I Service Water System. ) 1a 'b d 211.86'.-13 AF=T'T 12

o D " lD *D ~ Mf S n~%.I2Kw 2. From 3500F to 2000F (refer to FSAR Figur.cs 'i.'s.-4, 9.2 -4, and 9.2-1, 2 and 3) Q a. RHR Suction Isolat. ion Valves - 8701A and 8702A (to RHR Pump 1) n-

== E3 and 8701B and 8702B (to RHR Pump 2) - The two valves in each RilR

• m-subsystem are cach -powered from different er rgency power

_ d ' trains.' Failure ot either power train could prevent initisb on ~ %+p. of RHE cooling in the normal mannce from the contr.ol a (oom. Jn the event of such a failure, the affected. valve can be de-l 14 , energized and opened with its handwheel._J'Any other single fail-0 ++ ' ure can be toletated as it would only affect one of tha RHR 2 subsystems, and adequat.e cooling <an be provided by the oU ty redundant subsystem.

• +r 7 4 b.

RHR Pumps 1 and 2 4 Each pump.is powered from a different = 7e emergency power train. In the event of a single failure, either ~a 47 pump can provide sufficient FHR flow. 50 t I [~. c. RHR Heat 'Exchangers.l.dnd 2 . If either.he-at' exchanger is (- w-e .g x. una mi. table for, any reason., the Inmaining. heat exchanger can ~ , provide sufficient-: heat removal. capability. C Tw .O - *.d. -RHR. Flow Control Valves liCv603A and B - It.either mb these norm- %'%h ally open fail open valves clos.es spuriously, sufficient RHR a t.p cooling can be provided by the unaffected RHR train. b; a t e. RHR/ SIS Cold Leg Isolation Valves 8888A and B - If either of ,0, 44 s o* these normally open, motor operated valves, which are powered ,.} .+ A from different emergency power trains, closes spuriously, sufft-3 } cient RHR cooling can be provided.by the unaffected RHK train. 1 The affected valve can be deenergized and opened with its hand-N wheel. 3 [ 'l f. Component Cooling Water System - Two redundant subsystems are provided for safety related loads. Either subsystem can provide suf ficient heat-removal via one of the RHR heat exchangers. } s 1

m_m,

.v.--- v

b D*]D Dj M)b{ & cdJ o l[\\l, h lfbda g. Service Water System - Two redundant subsystems are pcovideo for .. }, safety related loads. Either subsystem can provide sufficient heat removal via one of the CCW heat exchangers. 7- ~ III. pration and Makeup (Refer to FSAR Figures 5.1-1,.6.3-1 and 9.3-16) d 1. Boric' Acid Tanks 1 and 2 Two boric acid tanks are provided, t.ach tank contains sufficient 4% boric acid to borate the' reactor cociant

system for cold shutdown.

i .y ~ 2. Boric Acid Transfer Pumps 1 and 2 - Each pump is powered from a 4 different emergency. power train. In the event of a single failure, either pump can provide sufficient boric acid flow. 3. Isolation Valve 8104 - If valve 8104, which is supplied from emer-gency power and is normally closed, cannot be opened due to powet train or operator failure, it can be opened locally with its hand-whee 1~. If valve 8104 cannot be opened with its handwheel, an 2, alternate flow path is available via a) air operated, fail open valve FCVc-113A and normally closed manual valve 8439, or b) gravity _ feed through r.ormally closed manual valves 8329 and 8331. n Refueling Water Storage Tank Isolation. Valves LCV115B oud LCV115D - ,4. ' Each valve is powered from a dif ferent emergency power train, only one of these normally closed motor operated valves needs to be l J; opened to provide a makeup flow path from the RWST to the Centrifugal Charging Pumps. L .s ~) 5. Centrifugal Charging Pumps 1, 2 and 3 - Pumps 1 and ' are powered from a different emergency. power train. In the event' of a single failure,- any one pump can pcovide suf ficient boration or makeup flow. ) 6. Charging Pump Suction Isolation Valves 8130 A, B and 8131 A, B - If l one of these normally open, motor operated valves, each of which is powered from a different emergency power train, closes spuriously, p operator' action can be used to deenergize the valve operator and ] reopen the valve with its handwheel. 3 e 211.86-15 AMENDMENT 12 FEBRUARY, 1979 I I

w 4 D M ]D rlo W3f,n W9JB oJhM >3 I 7. Normal Chargin; Fiow Control.Valv6 FCV-122 - This normally opeu "alve fails. op6n.on loss of air:6c: power. If FCV-122 closes sjuriously, the charging pumps can operate on their:miniflow cir- ,; ~ csits until operator:acti6p.can open bypass valve 8403. ~ ..,): 8. Nom's1 Char gin'g : Is ola ti6n : Va lve s L. 810T and: 8106 If either of these:uorcrally opeg motorloperal e4. valves, each of which is pwered from a different emergency-power < traitn, closes spuribusly, operator -action tan bec.used tor deeneigizerthe _ valve operatorJand reopen t.he 4 valve with its handwheel. ~. 9. Normal Charging-laolati6n-V Alve'.8146 -* If' the normally open valve closes spuriously, altergate'.chari;ing ' valve:8147 which fails open, can be used.

10. Charging Pum'p Discharg'e :Isoiali6n Va10es"8132 A, B hnd 8133 A, B -

If one of these normally openi motorloperated valves, each of which is powered from a different emergeni.y. power' train, closes -) spuriously,. operatur action can.tanus;;d t.o decoergize the valve, operator.~and reopen the~ valve with its handwheel. ~ I 11.' Reactor C6ol' ant Pump' Seal'Injectio'riaIsolati6n Viive' 8105 - If this nomally open, motut operated valve' closes spuriously, operator action can.be usell' to"deensigizer tne valve operator: and reopen the valve with its handwheel.

12. Reactor Coolant Pump Seal Injection Flow' Control. Valve HCV-186 -

~ ] This normally open valve. fails open on loss of-air'o,r power. If HCV-186 closes spuriously, the charging pumps can operate on their miniflow. circuits until:6perator action can open-bypass valve 8389. ia . - ;1

13. Reactor Coolant Pump' Seal Injection Valves 3102-AifB:and C - If any

? i of these normally open, motor operated valvos closes spuriously, operator action can be used to deenergize.the valve: operator and reopen'the valve with its handwheel. T,g S?;.k k. ) ,,[u

r DfD ~ %~ N QC uNG J bb$$$

14. Boron Injection Tank' Isolation valves 8803.A and B Dach valve is powered from a dif ferenc emergency power trains only one of these normally closed, motot operated val.ves needs to be opened to provide an alternate path 'and cource.- f.or bor a tion.

15. Borofi injection Tank Imlatiun h19t.s 8801 A'end B -- f ach valve is power (d f rom e dif ferent-emett,ency pawer train; only.one of thect be op ned to pra<ide p

norec11y cloced ndor ?>ocrated u.tves needs 19 3 an alternate path sad ' source 4.or boratson. & Add.Tusm f 3 .] J.V. Depr; 1surization (Refer to FSAR Figure 9 3-16) 1. Auxiliary Spray Valve 8145 - This normn11y closed valve fails closed on loss of air'or power. In this case, 8145 c'an be upened by us'ng -a. portable nitrogen bott:1e. -If 8J 4S I.s stuck closed as a result of a single failure, the redundant Seismic -Jategory.1 pressurizer safety valves cari he used to depressurize the RCS as: described % the alternate method for depressurization, 2' Charging Valves 8146.cnd 8147 i Thest vc1ves..fnit ope.n on loss at air or power. In this case, 8146 and 814/ can ce closed by using a portable nitrogen bo'ttle. If either as stuck.open,.tne redundant Seismic. Category 1 pressurizer safety valvei. can be used to depres-surize the RCS by venting the pressitrizer to the PRT (as de s ' r it,e d 9 in the alternate method for depressurization). 3. RHR Suction Isolation Valve 8701 A and B and 8702 A and B - The RHR 3 suction isolation valves are qualified for the steam line break I environment. Therefore, they are qualified for the less severe environment that would result if, as described in the:Above 1 and 2, y3 the RCS is depressurized as described in the alternate method for , -4 depressurization. ~ held Et 5uT C J l l l xy)A i ^ :. A AT:i"ENT 13 p, ec_,,

8 D**D 9 D 'II'D ' V ilIl V. Instrumentation E 4' Sufficient ' inst'rumentatioh iu provided to m6nitor from 'the Control Room i-the key function groocioted ii,th cold shutdown. All necessary indica-- tions are redtmdant2: thus;.1-4 the 'eeant of a single ~ failure, the opera-tor can make ' comparisons betecen duplicate information channels or 1 - } J betwee'n functiongily vdiated ch' ann'els in order to ideut'ify the partic.a- / ' lad ma'1 rune.tivtG 1eTer.h /wnt %: tion '/.5 -(Safety Related Display - lus t.rumen ta t' ion): 't onbTrpp1'itabl e details. }l TL.\\ R QualI$tkaTioN ~ The equipment. discussed in the cold shutdown scenario is safety grada. with thti :.01'owing :c tarificatica / 4h'ese are fen in number and of such a nature that. local manual' actions and/or equipment repair could be per~ formed while the pla'nt is: maintained in the hot standby. condition while preparations are" made to gd to coldMshutdown.: Thes'e manual acticns would tot' prevent tho p1' ant troa. achieving residual' heat removal system initiations.within 36 hours. i a Steam Ger arator power. ?opernt'ed relief valves. 1. 39 These air operated valve;( ace 'provlded with safety graala. renote opera- - - -c -tors, however remote control"pr/ visions are" control grade. Ilot l standb'y can be achieve'd a'nd i:iaiutained i> sing the sgf ety grade steam generator safety valves. The steam generator pbwez' operatyd relief e valves are provided with handwheels which can be o'p'erated locally to permit ~ plant cooldown. 1 I i 2. Charging line isolation valves,(8146, 8147)> a Scal. injection line Charging line auxill'ary spray valve (8145).m m hand control valve (HCV-186). Charging line flow control valve '(FCV-122). Reactor makeup control system flow control valve = 3 3 (FCV-113A). e D ~/ 211.86-18 3, AMENDMENT 19 } ~ JUNE, 1980 +

I -c, D"D D ' T hf f bu Wo u Tnese air ~ operated valves are not provided with safety grade remote operators, air supplies, or power supplies'. Their failure conse- ^ quences were discussed in the Parts.III and IV of the ::1f J.4 L.. -l _.,. jj gjpg y';j Q. ~ ? . = f,

• ~;

6 $, ',,):. 1 Residual". heat removal system flow conttol valve:(HCV-603 A/L) - This air opetated valve is not provided with e safety gradt-remote j' - operator, air supply'or power supply. Its failure consequence was S t uff/t f*//vre j discussed in Part II of the ~'A "r*>-- -~~e- ^ tValu=rs*eu, 4. Pressurizer Relief Tank This tank is a non-nuclear safety class and non-seismic category I tank. Its failure does not affect the ability'of the Virgil C. Sumer Nuclear Plant to achieve cold shutdown. AJJ Ig r. r S s .a $f a e*', s * ..s .. ^.. _J s I.. ' 211.86-19 . s:. ? AMENDMENT 19 .)j JUNE, 1980 s e

NATURAL CIRCULATION 3 The Virgi] C. Summer Nuclear Station and Diablo Canyon Unit 1 have been compared in det.11 to ascertain any differences between the two plants that could potentially affect natural circulation flow and attendant boron mixing. The general configuration of the piping and components in each reactor coolant loop is the came in both the Virgil C. Summer Nuclear Station and Diablo Canyon, Both plants have Model 93A reactor coolant pumps. The Virgil ] C. Summer Nuclear Station has Model D3 steam generators and Diablo Canyon has Model 51 steam generators. The elevation head and ficw resistances re-presented by these components and the systen piping is similar. To compare the natural circulation capabilities of the Virgil C. Summer Nuclear Station and Diablo Canyon, the hydraulic resistance coefficients were compared. The coefficients were generated on a per loop basis to permit such a comparison between a three loop and a four loop plant. The hydraulic resistance coefficients applicable to normal flow conditions are as follows: Virgil C. Summer Diablo Canyon Unit 1 Nuclear Station Reactor Core & Internals /.6 x 10-10 ft/(loop gpm)2 Reactor Nozzles 36.8 (laterl RCS piping 24.0 Steam Generator 114.4 182.8 P l/2 Flow Ratio Virgil C. Summer =.182.8 = (later) Diablo Canyon (later) D The general arrangement of the reactor core and internals is the same in the Virgil C. Summer Nuclear Station and Diablo Canyon. The coefficients in-3 ~ dicated represent the resistance seen by the flow in one loep. As exhibited, the difference between the internals of a three loop and a four loop plant results in a higher coefficient for the Virgil C. Summer Nuclear Station. The reactor vessel outlet nozzel configuration for both plants is the same. The radius of curvature between the vessel inlet nozzle and down-comer section bf the vessel on the two plants is different. Based on 1/7 scale model testing performed by Westinghouse and other literature, the f radius on the vessr1 nozzic/ vessel. downcomer juncture influences the hydraulic resistance of the fic turning from the nozzle to the downcomer. 7 The Diablo Canyon vessel inlet nozzle radius is significantly smaller than that of the Virgil C. Summer Nuclear Station, as reflected by the higher co-efficient for Diablo Cr.pyun. S

The resistance (coefficient) for the RCS piping for both plants is the same. Details of the specific steam generator units were also compared to as-certain any variation (e.g., primary volume, tube height, tube diameter) that could affect natural circulation capability by changing the effective eleva-tion of the heat sink or the hydraulic resistance seen by the primary coolant. It was concluded that there are no differences in'the original design ot the steam generators in the'two plants that would affect the. natural circulation characteristics. As indicated, the difference between the total resistance coefficients O for the two plants is insignificant. It is expected that the telative effect of the coefficients would be the same under natural circulation conditions cuch that the natural circulation loop flowrate for the Virgil C. Summer Nuc1 car Station would be similar to that for Diablo Canyon. The coefficients provided reflect the flowrate and associated heat removal capability of an individual loop in the plant. The comparison, therefore, does not take into consideration the number of loops available nor the core heat to be removed. An evaluation of the Virgil C. Summer Nuclear Station Ibin Steam and Emergency Feedwater Systems has been performed to demonstrate that cooling can be provided via two steam generators following the most limiting single active failure, J.e., the failure of a steam generator 7 power operated relief valve. Loop natural circulation flow is dependent on reactor core decay heat which is a function of time based on core power operating history. Under natural circulation flow conditions, flow into the upper h'ead area will con-stitute only a small percentage of the total core natu'ral circulation flow b: and therefore will not resuit in an unacceptable' thermal / hydraulic impedance to the natural circulation flow required to cool the core, For typical 3-loop and 4-loop plants (including the Virgil C. Summer i Nuclear Station & Diablo Canyon) there are two potential flow paths by which flow crosses the upper head region boundary in a reactor. These paths are the head cooling spray nozzles and the guide tubes. The head cooling spray 7 nozzles constitute a flow path between the downcomer region and the upper head region. The temperature of the flow which enters the head via this path 4 corresponds to the cold leg value (i.e. Teold). Fluid may also be exchanged between the upper plenum region (i.e., the portion of the reactor between ~ the upper core plate and the upper support plate) and the upper head region via the guide tubes. Guide tubes are dispersed in the upper plenum region from the center to the periphery. Because of the nonuniform pressure distri-bution at the upper core plate elevation and the flow distribution in the upper plenum region, the pressure in the guide tube varies from location to location. These guide tube pressure variations create the potential for flow to either enter or exit the upper head region via the guide tubes. To ascertain any difference between the upper head cooling capabilities between Diablo Canyon and the Virgil C. Summer Nuclear Station, a comparison of the hydraulic resistance of the upper head regions were made. These flow paths were considered in parallel to obtain the following results. i

I Diablo Canyon Virgil C. Summer Unit 1 Nuclear Station Flow Area (ft2) 0.77 1.51 Loss Coefficient v Overall Hydraulic Resistance 2.57 (ft-4) Relative Head Region Flowrate 1.00 (later) (Based on Hydraulic Resistance)- Head Region Flow Rate Relative 1.00 to Loop Flow As indicated above the effective hydraulic resistance to flow in the ] Virgil C. Su=mer Nuclear Station is (*) times greater than Diablo Canyon. As-suming that the same pressure differential existed in both plants the Virgil C. Summer head flow rate would be 90 percent of the Diabla Canyon flow. Virgil C. Summer is a 3-loop plant and Diablo Canyon is 4-loop; therefore, in terms of relative portions of loop flow communicating with the head region, the Virgil C. Summer head flow as a fraction of loop flow is (*) percent greater than the corresponding Diablo Canyon fraction. In addition the overall mass of metal associated with the Virgil C. Summer Nuclear Station upper head is significantly less than for Diablo Canyon due.to the smaller physical size. Thus the upper head cooling capability at the Virgil C. Summer Nuclear Station would be no worse and would likely be better than demonstrated by the Diablo Canyon natural circulation cooldown test. It can, therefore, be concluded that the results of the natural circulation cooldown tests performed at Diablo Canyon will be representative of the natural circulation and boron mixing capability of the Virgil C. Summer Nuclear Station. The results of these tests will be reviewed for applicability. A natural circulation cooldown test will be performed at the Virgil C. Summer Nuclear 6 Station if the Ciablo Canyon prototype test, or similar test at another nuclear plant, is not completed or does not provide satisfactory results during the first fuel cycle at the Virgil C. Summer Nuclear Station. 3 v ~ r P P

• (later) a I'

'l I l

s INSERT A TECHNICAL REQUIREMENTS g The safe shutdown design basis of the Virgil C. Summer Nuclear Station is hot standby. Under abnormal conditions, the plant is designed to remain in a safe hot standby condition until (a) normal systems can be rectored to permit either return to power operation or cooldown to cold shutdown conditions, or (b) sufficient systems capability can be restored (depending on plant condi-tion) to permit cooldown to cold shutdown conditions under abnormal plant conditions. This design basis is considered to constitute a safe design. O BTP RS3 5-1 establishes specific design requirements that address the various system functions that are seguired to achieve and maintain a safe hot standby and cold shutdown condition. BTP RSB 5-1 requires plants with construction permits docketed after January 1, 1978, to comply in full with the design requirements of the BTP. Plants with construction permits docketed prior to January 1, 1978, (including Virgil C. Summer) are required to address the BTP technical requirements and demonstrate partial compliance. The following is a discussion of the Virgil C. Summer Nuclear Station compliance with the technical requirements of BTP RSB 5-1. This discussion demonstrates that under the postulated condition of BTP RSB 5-1, that the Virgil C. Summer Nuclear Station can be maintained in a safe hot standby condition and taken to Residual Heat Removal System -(RHRS) initiation con-ditions within approximately 36 hours, including credit for limited manual actions outside the control room to operate and/or repair a limited number of components that are not safety-grade or single failure proof. Itemized below are the technical requirements of BTP RSB 5-1 followed by a general discussion of the Virgil C. Summer Nuclear Station compliance. The Technical Requirements section of this response is then followed by more detailed discussion in sections entitled Cold Shutdown Scenaro, Single Failure Evaluation and Natural Circulation. g N d r, k

INSERT B I Part of the low-power testing program at the Virgil C. Summer Nuclear Sta-tion includes several tests to verify natural circulation. However, this test program does not include boration or cooldown to RilRS initiation conditions. Diablo Canyon has committed to perform a natural circulation boration and cooldown' test in compliance with BTP RSB 5-1 requirements. Based or. a com-parison of the Virgil C. Summer Nuclear Station and Diablo Canyon Unit 1 natural circulation capability (see natural circulation evaluation provided below), it has been determined that the natural circulation cooldown tests to Q be performed at Diablo Canyon will be representative of the natural circula-tion cooldown and boron mixing capability at the Virgil C. Summer Nuclear Station. The results of the testing at Diablo Canyon will be reviewed and a natural circulation cooldown test will be performed at the Virgil C. Summer Nuclear Station prior to startup following the first refueling if the Diablo Canyon prototype test, or similar test at another nuclear plant, has not been completed or does not provide satisfactory results. The results of this testing will be used to confirm cooldown times, cmer-gency feedwater requirements and operating procedures. I, b t ~. l'. l~ It 1 (; l. I!'; mn--,,-e,.-,,..-.. .----,_,.,_n-.. -,-._,n-,,,-,,.,,;a,,-n-.,,-_, -,,,,,.,,,,.,,,.,.,,n,,n- ,,,,e m n,. ,,,.e,,,,,.,

i INSERT C l Alternative methods of depressurization, after boration and RCS cooldown to 4500F, consist of (1) discharging reactor coolant from the pressurizer to the pressurizer relief tank via the pressurizer power operated relief valves, and (2) allowing the pressurizer to cool via ambient head losses as the teactor coolant system is maintained at 3500F via natural circulation. The former alternative provides relatively rapid depressurization capability to permit timely RllRS initiation and EFS termination. The latter alternative permits a gradual depressurization but requires extended EFS operation to remove core decay heat to maintain the RCS at 350 F while the pressurizer cools via ambient heat losses. Depending on plant conditions, either of these two alternatives may constitute the preferred ceans for plant depressurization. p i o J-4

i INSERT D The cold shutdown scenario of maintaining RCS temperature and pressure' without letdown is considered the limiting cold shutdown scenario and is I presented to demonstrate cold shutdown capability under abnormal conditions. As additional. components are assumed to be available, the cold shutdown acenario is simplified. For exampici as the reactor vessci head vent valves or the pressurizer PORVs are assumed to be available, the boration and makeup function or the depressurization function respectively, is simplified as RCS letdown becomes availabic. 3 l i, 1 W i l N -Y 4 l t

4 4 4 INSERT E The results of the calculations described above demonstrate that, based ~ on the assumed initial conditions, boration and depressurization with 4 wt % boric acid can be accomplished without letdown and without taking full credit for the available volume created by the.cooldown contraction. However, the operator may elect to borate using the 12 wt % boric acid contents of the boron. injection tank as well as 4 wt % boric acid from the horic acid tanks. Should boration without letdown prove impractical due to any combination of plant conditions or equipment failures, letdown can be achieved by discharging RCS' inventory via the pressurizer power. operated : elief valves or the reactor Q vessel head vent valves. 9 s 4 p J W

1 INSERT F 16. Reactor Vessel Head Vent Valves 8095 AB and 8096 AB 2 These normally closed valves fail closed on loss of power. They are ar-ranged with two valves in series in each of two parallel paths from the reactor vessel head. The series valves in each parallel path are powered from the same emergency electrical power supply and the two paths are powered from separate emergency power cupplies. This valving and power supply arrangement ensures that one path from the reactor vessel head a can be opened assuming a single failure. One' path is sufficient to permit letdown from the reactor coolant system to augment boration and makeup operations. p l P 0 9 G I e 5 a o

L INSERT G 4. Pressurizer Power Operated Relief Valves PCV-444B and PCV-445A These normally closed valves fail closed on loss of air or power. How-ever, the valves are redundant, are powered by separate emergency electrical power supplies and have backup scismic category nitrogen supply accumulators. The operability of either valve is sufficient to permit depressurization. ] h* w M Ie 4

.-_ - -...= ~ - i t INSERT H 4 S. Pressurizer Power Operated Relief Valves j These air operated valves are not supplied with safety grade remote operators, however they are powered from separate vital AC electtical power supplies and two of them have a seismic categary 1 nitrogen supply accumulators. These valves constitute one of several depressurization alternatives as discussed in Part IV of the cold shutdown scenario . Refer to the response to Question 211.10 for a discussion of the j operability of these valves relative to RCS cold overpressure mit;igation, i i t 3 i '( I 's 1 s L 1 A m .}}