ML17332A753

Proposed Tech Specs,Incorporating 2.0 Volt Interim SG Tube Support Plate Plugging Criterion for Fuel Cycle 15

ML17332A753
Person / Time
Site:	Cook
Issue date:	04/25/1995
From:	INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:
Shared Package
ML17332A752	List: ML17332A751 ML17332A753
References
NUDOCS 9505010275
Download: ML17332A753 (37)

Text

ATTACHMENT 2 TO AEP:NRC:1166R EXISTING TECHNICAL SPECIFICATION PAGES MARKED TO REFLECT PROPOSED CHANGES 9505010275 950425 PDR ADQCK 05000315 P

PDR

1

REACTOR COOLANT SYSTEM STEAH GEM:RATORS XNiG CONDITION FOR OPERATXON pj,Q(~ Ag CH.PI c E.0/JL.'(

3.4.5 Each steam generator shall be OP~LE.

APPLICABILITY:

MODES 1, 2, 3 and 4.*

ACTION:

With one or more steam generators inoperable, restore the inoperable generator(s) to OPERABLE status prior to increasing T,~ above 200 P.

0 SURVEXLLANC RE..

S 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the follo~ing augmented'nsexvice inspection program and the requirement ox Specification 4.0.5.

4.4.5.1 Steam Generator Sa e Se>ection and Inspect on - Each steam generator'hall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number ox steam generators specified in Table

4. 4-1.

4.4.5.2 Steam Gene ator be S

le Selec ion and Inspection - The steam genexator tube minimum sample si"e, inspection result classification, and the cox espond~g action requi ed shall be as specified in Table 4.4-2.

The insezvice inspection ox steam genex'ator tubes shaLl be performed at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shaU.

be vez~~ied acceptable pex'he acceptance c iteria of Spec~&ication 4.4.5.4.

The

.abes selected xor each fnsexvice ~inspection shall include at leas" 3'4 ox the total nu ber ox tubes in all steam gene atozs; the tubes selected for these inspections shall be selec ed on a random basis except:

Where experience in s~ar plants vith similar vater chemistry indicates c itical areas to be inspected, then at least 50'4 of the tubes inspected shall be from these c it'cal azeas.

b.

The first sample of tubes selected for each inserv'ce inspection (subsequent to the preservice inspection) ox each steam generator shall include:

l.

All tubes that previously had detectable vaU. penetrat'ons (greatex than or equal to 20'4) that have not been pt.ugged or repai=ed by sleevtng i6 the af ected area..

Soec'='cat'on as Long as L'ainta'wed does uot apply La '.Lode 4 vhile pe =oring crevice flushing Condit'ons for Operat'on for Specification 3.4.1.3 are COOK NUC~KR PLANT UNIT L 3/4 4-7

REACTOR COOLANT SYSTEM SURVEILLANCE REOUH~EHTS 2.

Tubes i.n chose areas where experience has indicated pocencial problems.

3.

A cube inspection (pursuanc to Specification

4. 4. 5.4. a. 8) shall be performed on each selected tube.

If any selected cube does noc permi.c the passage of tha eddy current probe for

~X~ e 0h a cube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a cube inspection.

c.

In addicion to tha sample required in 4.4.5.2.b.l chrough 3, all tubes vhich have had che P+ criteria applied vill be inspected in che roll expanded region.

The roll expanded region of these cubes may be excluded from ehe requiremenes of 4.4.5.2.b.l.

d.

The tubes selected as the second and chird samples (if required by Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:

The cubes selected for ehe samples include the tubes from those areas of ehe cube sheec array where cubes vich imperfect ons vere previously found.

2.

The" inspections include those portions of the tubes vher e imperfections vere previously found.

esupp~~Mtm The results of each sample inspection shall be classified inca one of the following three categories:

~Cave ezv Ins action esu ts C-1 Less than 5X of che cocal cubes inspected are degraded tubes and none of ehe inspected tubes are defective.

C-2 One or mora

tubes, buc noc mora than 1X of the tocal cubes inspected are defective, or becveen 5X and lOX of the tocal cubes inspeccad are degraded cubes.

C-3 Mora than 10X of.the tocal tubes inspecced are degraded tubes or mote than 1X of the inspected tubes a

a defective.

uc IDYL't

'fO

J

b.

4.

Tubes left in service as a result of application of the tube support plate plugging criteria shall be inspected by bobbin coil probe during all future refueling outages.

INSERT "B" e.

Implementation of. the steam generator tube/tube support plate plugging criteria for one fuel cycle (cycle 15) requires a 100 percent bobbin coi 1 inspection for hot-leg tube support plate intersections and cold-leg intersections down to the lowest cold-leg tube support plate with known outside diameter stress'orrosion cracking (ODSCC) indications.

The determination of tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20 percent random sampling of tubes inspected over their full length.

REACTOR COOLANT SYS SGR XLLANCZ RE UTM KHTS Continued Hate:

In all inspections, previously degraded tubes must exhibit significant (greater than oz equal to 104) further waLL penetrations to be included in the above percentage calcuLations.

4.4.5.3 Insaection Fre encXes - Tha abave zequi"ed inseWce inspections of steam generator tubes shall be performed at the following f aquencies:

a.

The first insezvice inspection shall be performed after 6 Ef"ective Full Power Hanths but within 24 calendar months of initial criticality.

Subsequent insezvt.ca inspections shall be per armed at intervals of not lass than 12 nor more than 24 calendaz months arte the previous inspection.

If two cansecutive inspections follawing service under AVT conditions, not including the preservica inspection, result in all inspection results faLLing into the C-1 category or if two consecutive inspections demons=ate that previously observed degradation has not continued and no additionaL degradation has occu ed, the inspection intaxval may be extended to a maximum ox once pe 40 months.

b.

Xf the results oz insezvica inspection of a steam genezataz conducted in accardanca with Table 4.4'-2 at 40 month intexvals fall in Category C-3, tha inspection frequency shall be incxeased ta at least once per 20 manths.

The increase in inspection fxequency shalL apply tnt'. the subsequent inspections sat'sfy the criteria of Specification 4.4.5.3.a; the interval may then be a~ended to a maximum of once pez 40 months.

C ~

Additional, unscheduled inservica inspect'ons shall be performed on each steam generator in accordance with the first sample inspection specified h Table +.4-2 8u ~zg the shutdawn. subsequent to any or the foLlowing conditions:

L.

Primary-ta>>secondary tubes Leaks (not includ'sg leaks originating from tube-to-tube sheet welds) in excess of the Limits of Specification 3.4.6'.2.

2.

A seismic ace~

ence graataz than the Qperat'ng Basis Earthquake.

3 ~

A loss-of-coalant accident requiring ac sation of the engineezad saxeguards.

oea~e'I 4.

A main steam lm~e ox feedwater line break.

I

%~asm-re ui~&appL-MMkonw&eh~b~upaorr ply-4nc~~eggng-e fata ia-shaLL-b&xspected-by-babb in - coil p5fA A hh COOK HQCL:-UR. P~IT - UViI 1

3/4 4-9 A?

HD~HT VO.

04

1

~\\

M

~

vA~R COOLANT SURVEZL~C RK U ~Z'~NTS Continued 4.4.5.4 coco ance Crite

~a a.

As used in this Specification:

Im

<<fec"'on means an exception to the dimensionsf finish o contour of a tube or sleeve from that raqu3 "ed by fabri'cation drawings o" spec'ficat'ons.

Eddy-cur=ant tasting indicat'ons belo~ 20% of the ncminal tube wall thicknessf if detac"able, may be consida ad as i~fee=iona.

2 e gene al corrosion ocax=ing on ei Ne inside or outside of a tube or sleeve.

3 e De aded Tube oz Sleeve means an imps faction g eater than or equal to 20%

of the nominal wall thickness caused by deg=adat'n.

4 ~

Pe>> ent De adation means the amount of the t"he wall thickness af ac=ed or removed by degradat'on.

Defac means sn imnec ec 'an af sech seve 'af that it exceeds ths egai limit.

6 Reoai= P'u in Limet means the imperfec=ion depth at or beyond which the tube or sleeved tube shall be epa'"ed or removed fram service.

Any tube which, upon inspec ion, exhibits tahe wall degradation of 40 percent or more of the nominal tube wall thickness shg.l be plugged or repai ed pr'or to returning the steam gene ator to service.

This definit'on does not apply to the por"ion of the tube in the tuhesheet below tha F< distance for F+ tubas.

Any sleeve which, upon inspec=ion, exhibits wall dag adat'on of 29 pe cant or mora of the nominal wall thickness shall'e plugged pr'or to returning the steam generator to service.

Zn addi 'on, any sleave exhibiting any maasu able wall l

in sleeve expansion transition or weld cones shall he plugge T

oua~&t5

<<s V

7 ~

~

<<s<<>>

~~uppo rt~Xwue e Unse~iceable desc=ibes the cond't'on of a tube or sleeve 'f leaks or ccnta'".s a defac= large enough to af ec" its struc=ural intag='ty in the even-of an Operating Basis Ear hquake, a loss-of-coolant acc'dent, or a steam line or feedwater line break as spec': ed in 4. 4. 5.3. c, above.

8.

~nenes

'an detaw'nes tha cend't'cn af the steam gene"stn tahe or sleeve from the point of ante (hot leg side) completely C00K NUC AR PMT UNX l 3/4 4-l0

~NDF:-NT NOe

REACTOR COOLANT SYST~"f

'URVRI~~CE REOUIR~ViS (Continued) around the U-bend to the top support of the co d leg.

For tube in which the tube suppor" place elevac'on entarim plugging limi has been applied, the inspect on will inc uda all the hoe leg intersect'ons and all cold leg intersect'ons down to, at

least, the level of the last crack indication.

9.

~SLeevin a enhe is paretic e-d only in areas where ehe sieeve spans ehe tubesheet area and vhosa lover joinc is at the primary fluid tubesheet face.

10.

l&eleth~

The Tube Suoocrt Plate nea~

m P Lu i ~

Crite<ia

's used foz'.

e for continued service that>4s ted stress corrosion cracking e

cube suppor" plates

~

For te interim plugg'ng limS.t, the ce will be based upon standard ant-specific gu'delines used for appropriate to accommodate the uata tube suppor" plate signals e/depth parameters.

Pending fication requir ments i.n ASHE andard cafibrat d against the in the Donald C.

Cook Nuclear ccions 'for consistent voltage

'the si al amo itude of a crack 1 to 2.0 volt, zagardless of the naif, as a resul",

the projected c ack indi.cations is verified to leakage less than 12.6 gpm in the tad steam line br ak event.

The expectecb leak "ates from 'the ust be consistent -ith 'iCAP-13187, dra=t NUREG-1477.

paired if the si>ml amplitude of r than 2.0 volt ex"apc as noted in ith a bobbin coil s'gnal amplicude s than or equal to 3.6 volts, if a tion does not deeact deg adation.

th a bobbin coil s'gnal amplitt de e plugged or repa'"ad.

indicat on i.s less than or e depch of cube vali. penecraci end-oz-cycle distribution o

result in primary-to-seconda faulted loop during,.a pos a

mechodo logy for calculacin projected crack d~seribution Rev.

0, and as prescribed i A tube should be plugged or e

the crack'ndicati.on is grea a

4.4.5 P.a.10.3 below.

A ~be can zemain in service greater than 2.0 volt but le rotat ng pancake probe inspe Indicat ons of deg adat'on greater than 3.6 volts vill 5 2.

3.

de osicion of a steam generaeor eub expe ienc'ng outer diameter ini ia confin d within the thickness o -- th applica on of the cube support la cube's dis osition for concinued bobbin pro'oe qignal amplitude.

Th pl all inspections shall be amended as additional infohnation needed to al with respect to ~the above vo tag i.ncorporacion of tha voltage v ri standard ve ificationa an ASME st laboratory standard vilh<be util ed Plane Unit 1

steam genez'ator i pe nozmalization.

1.

A tube can remain in servi.ca if'

-'oward the bot"om of the eubeshee t thac has bee..

i conse

~at ately deca~

ned =o be 1.1 inches (noc includ'ng eddy outran= unce =aint: ).

12.

F+

hube 's a "uoe 3raae

<<nan 40se

~

wit¹n the F>> d's wi h deg adat=on.

be'ow tha F>> d stance, equal -o or and not deg"aded (i.e.,

no ind catons oz crack-ng) rance.

COOK XC:"--'3. Py 3/4 4-L'"=".D<=':

NG. :

INSERT "C" This definition does not apply to tube support plate intersections for which the voltage-based plugging criteria are being applied.

Refer to 4.4.5.4.a. 10 for the plugging limit applicable to these intersections.

INSERT "D" 10.

Tube Su ort Plate Re air Limit is used for the disposition of a steam generator tube for continued service that is experiencing outside diameter stress corrosion cracking (ODSCC) confined within the thickness of the tube support plates.

At tube support plate intersections, the repair limit is based on maintaining steam generator tube serviceability as described below:

a 0 b.

c ~

Degradation attributed to ODSCC within the bounds of the tube support plate with bobbin voltage less than or equal to 2.0 volts will be allowed to remain in service.

Degradation attributed to ODSCC within the bounds of the tube support plate wi th a bobbin voltage greater than 2.0 volts will be repaired or plugged except as noted in 4.4.5.4.a. 10.c below, Indications of potential degradation attributed to ODSCC within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts but less than or equal to 5.6 volts may remain in service if a rotating pancake coil inspection does not detect degradation.

Indications of ODSCC degradation with a bobbin voltage greater than 5.6 volts will be plugged or repaired.

<AC OR COO~)r.

SvS

=wS SVRV

~ MC= RZ C".RZwvNTS (Continued'l

.".e stean gene ator shall be detained OP="BABL=- after complet'ng the corresponding actions (plugging or slaaving all tubas axcaading the repa'"

imit and all tubes containing th=ough-wall cracks) raqui ed by Ta"la 4.4-2.

C ~

S"oat gana ator tube repai s

may be made in accordance wi h tha methods described in either RCAP-12623 or KN-313-P.

4.4.S.S Re+orts Following each insezvica inspection of staam'generator tubas, if there a=e any tubas requiring plugging or sleeving, the numbe of tubas p'ugged or sleeved in each steam generator shall be reported to the Cc~ission within 15 days.

b.

The ccrplete results of the steam gene ator rube inservica inspection shall be incLuded in the Annual Operat'ng Rape= fo" tha per'od i,n wh'ch th's inspection was completed.

Th's apo shalL incLude:

Mbe and ex=ant of tubas inspected.

2.

Locat on and percent of wall-thickness penatrat'n fo" each indica 'on of an ~

ec"ion.

(~se~Y <

Co 3.

Zdantificat'on of tubes plugged or sleeved.

Resul s of s

aam generato tube inspect'ons wh'ch faLl into Category C-3 and rendu' prompt no"' icat'n of he Coctissi on shaLL be apor ad pursuant to Spec'cation 6.9.l pr'or to rasumpt'on of pLant ope ation.

The wr'tten followup of this ra o= shall provide a

descr'pt'on of investigat'ons conducted to dateaina cause of the tube degradat'on and cor"ac=ive measures taken to prevent acurrance.

d.

Tha Lts of inspec=iona per o wh'ch

"".e e

suppor=

plate a=-ed o" that dafac=s b

p'ged eha'L be sapor=a the the L,nsoect ion.

he e

s L'sting of aoo~ '

tubes ed under 4.4.5.2.'"" al'tubes in ter~

plugg'ng cr'er a

has beer.

ow tha '~ d'=anca and wa e

no" o=ai~s'cn witn'n 15 davs following e

~ j p

~

~

)

C dagradat'on (voltage).

cn (applxcab e

an amac ons er be an en o

%44 9e(h'ek ~~

~ 4

/S COOK NUC-:.AR P~~: - UN:

1 3(4 4-12 I

At - NDF:-N NO. VH, ~~)..)

l.78

INSERT "E" 4.4.5.5.

~Re orts d.

For implementation of the voltage-based repair criteria to tube support plate intersections, notify the staff prior to returning the steam generators to service should any of the following conditions arise:

If estimated leakage based on the actual measured end-of-cycle voltage distribution would have exceeded the leak limit (for the postulated main steam line break utilizing "Standard Review Plan

NUREG 0800" assumptions) during the previous operating cycle.

mew: IR 2.

3.

4 ~

If circumferential crack-like indications are detected at the tube support plate intersections.

If significant indications are identified that extend beyond the confines of the tube support plate.

If the calculated conditional burst pro/ability, as calculated per WCAP-14277, exceeds I X 10

, notify the NRC and provide an assessment of the safety significance of the occurrence.

3.4.6.2 Reactor Coolant Syseem leakage shall be limited to:

No PRESSURE BOUNDARY LEAKAGE, b.

1 GPH UNIDENTIFIED LEAKAGE, Ce 600 gamous per day eotal p

-eo-secondary leakage through all steam generators and 150 lons per day through any one steam generator for Fuel Cycle e.

10 GPM IDENTIFZED LEAKAGE from the Reactor Coolant System, Seal line resistance greater than or equal to 2.27E-l ft/gpm

and, The leakage from each Reactor Coolant System Pressure Isolation Valves specified in Table 3.4-0 shall be limited to 0,5 gpm per nominal inch of vaLve size up to a umdmum of 5 gpm, at a Reactor Coolant System average pressure within 20 psi of tha nominal full pressure value.

~CTZON:

Pith any PRESSURE BOUNDARY LEAKAGE, be in ae least HOT STANDBY.

within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Pit!x any. Reactor". Coolantt,System leakage greater than any ona of thy above

Hairs, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage rata to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in ae lease HOT STANDBY vithin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN wi.thin the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Ce Pith any reactor coolant system pressure isolaeion valve(s) leakage greaten than tbe abave linda, declare tba lealdng valve ixmyerable

[

and isolate the hi,gh pressure portion of the affected system from the low pressure portion by the use of.a combinaeion of ae least tvo closed valves, one of vhich may'e the OPERABLE check valve and the other a closed da-energized moeor operated valve.

Verif'y the isolated condition of the closed da-energized moeor operaead valve ae lease once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or be in at least HOT STANDBYwithin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the folloving 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Specif cat on 3.4.6.2.e is applicable with average pressure within 20 psi of the nominal &d.l pressure value.

COOK NUCLEM PLANT - UNIT 1 3/4 4-1.6 AH~MEH7 NO. 46Z-, 466-, ~, (88

RZAC. OR COOLANT SYSTZH AS=S 3 4.4.5 STENCH CFN.RATORS TUBE YN.=CRT~

Tha Suveillanca Requirements for inspec=ion of che steam generator tubes ensure that the stwccu al integrity of this por 'n of the RCS will be mainta'ned.

The program for inservice inspec"'on of steam gene acor tubes is baaed on a mod'icac'on o

Regulatory Cuida I.S3, Revision 1.

Insevice inspect'n of steam generator tubing is essanrial in order co main ain survelan'a of the condit'ons of the tubes in the event that the e is evidence of mechanical damage oz'rogressive degradation due to des'gn, manufaccu

'ng errors, or insevica conditions that lead to corrosion.

Znsavica inspec=ion of steam gene asar tubing also,p~ides a means of character'-ing the natu e and cause of any tube degradat'on so thar. cozzactive measure's can be taken.

The plant is expec ad to be ope aced in a manna such tha

. e second-ary coolant will be ma'ncained within those chemistry limits und to rasul in negligible corrosion of cha steam generator tubes.

Zf

. a secondary coolant chemistry is not main ainad within these parameter 1

ts, locali=ed corrosion may likely result in stress corrosion cracking e extan" of crack'ng du"'ng plant operat'on

~ould be 1'-'cad by the 1'

acion of steam gene ato" tube leakage between the primary coolant system d the secondary coolant system.

The al'wahe p imag-co-seconda=g leak

~

is 150 gallons pe day pe" steam generator for one fuel cycle (Cycle 1

}

Axial or cir~ferencially or'anted cracks having a

primary-co-second leakage less than this limit dur'g operat'on will have an adequate margin of safety to withstand the loads imposed during normal operac'on and by posculared acc'dents.

Leakage in excess of this lim'c will require plant shutdown and an inspection, during which the leaking tubes wi' he located and plugged or rapa3 "ed.

A steam gene acor while undergo'ng crevice flushing in Mode 4 is availahla for decay heat removal and is operable/ope acing upon ra'nscacemenc of auxiliary or main feed flow control and salaam con ol Wastage-type defects ara unlikely with the all volac'e treatment (AV. } of secondary coolan".

However, even 'f a defect of similar type shculd develop

sevica, i" wi' he found d.='ng scheduled inse

~ice steam gene"ator tube examinac'..s.

Plugg'ng o" sleeving will be rendu'ed for all tubes with imperfactions exceeding the repa' '-'

wh h is defined

'n Soec' 'ac'n 4.4.5. 4. a.

Steam gene ac'"z tube inspections of ope ac'g plants have demonstrated the capah'1'cy co reliably detect degradac'on

that, has penetrated 20%

o che or'ginal t he wall thickness.

= hes expa 'enc'ng outa" diameter stress corrosion crack'".g wi"h'n the thickness of the tube suppor p'ates aze plugged or repa'-ed bv the cz'ter'a of

4. 4. 5. 4. a. 10 COOK NUC::-AR P~i: -

UN B 3/4 4-Za AY=NO..')

NO. 444, Mrs )

~ 8

O OO (Canthmed)

%+never tha rasuLts o8 any steam generator tubing insane ca Qzspec fon ZaLL Quito Category C-3, these results wi11 be promprLy raportad to tha Commfssfon pu susnt to SpecfÃfcatfon 6.9.1 prfor to raenptfon of plant operatfon.

Such cases vill be consfderad by rha ComuCksfon on a casa-by-casa basfs and.may result fn 2 ra~ament for enRLysfs Laboratory dxxafnatfons, tests, addftfonaL eddy-ca ant fnspec fon, and rav&on of the Te~csL Spec f fc3tfons, jf nacessaryo COOK HUCL:AR F~

UHXi L 8 3/4 4-Zb

~~

HO 4"" 444- '

~ACTOR COOLANT SY BASKS A

Degraded sceam generacor tubes may be repaired by.ehe install'acion of sleeves which span ehe section af degraded steam generator tubing.

A sceam generator cube wi.eh a sleeve instaLLed meets the sutural requirements of tubes which are noc degraded.

To determine the basis for the sleeve plugging Limit, the minimum sleeve walk thickness was calculated in accozdance with Draft Regulaeozy Guide L.L21 (August 1976).

In addition, a cambined allawance of 20 percent of wall thickness is assumed for eddy cuz enc tescing inaccuracies and cancinued operational degradacian per Draft Regulatory Guide 1.121 (August L976).

The following sleeve designs have been found accepcable by the NRC staff:

1.

Vescinghouse Mechanical Sleeves

(%CAP-12623) 2.

Cambuscion Engineering Leak Tight Sleeves (CEH-313-P)

Descript ons oz other future sleeve designs shall be submitted to ehe NRC for review and approval in accordance with LOCZR50.90 prior to their use in the repair of degraded steam generator tubes.

The submietals related to ocher sleeve design shall be made at lease 90 days prior to use.

R3 ACTOR COOLANT SYS'AAA'AB 3

3 ~.~.6 RE23.CTOR COOLQPI'YST~e L~<XAGE 3 L.L.6

+%CAGE DET ~iON SYS~1c..S The RCS leakage detection systems required by this spec.ficatian are provided eo monieor and detec" Leakage f om che Reactor Coolant Pressure Boundary.

These detection systems are consistent with ehe recom=endacions of Regulatory Guide 1.45, "Reac or Coalanc P assure Boundary Leakage Detection

~.:stems,"

May 1973.

3 L L.6.

OP~AMONAL GE Indus~ exper ence has shown ehac while a limited amounc of leakage i.s expected f am ehe

RCS, ehe unidenc'f'ed portion of this leakage can be reduced eo a threshold value of less chan l gpm.

This threshold value is sufficiencly Low eo ensuze early detection of additionaL leakage.

The 10 GPH IDBKI."IEDI~AGE Limicat ons provides allowance for a Limieed amount of Leakage fram know sauzces whose presence wiLl not interfere with the dececcion oz UNIDBKIFZH L:-%AGE by the leakage decect'on systems.

The 1'tation an seal L'ne resistance ensures that ehe seal line resistance is greater than or equal to the resistance assumed i.n the minimum safeguazds LOCA ana1.ysis.

Th's analys s

assumes that a1.1. of the flaw thac is d ~creed from che boron inject. an Line co the seal injec" on Line is unavailable for care cooling.

COOK NUCT:-BAR. ?~~T

- UNIT L 3 3/A A-3

~.-Hi BO. P3. Q, g.

162 Car=ected Pap

PA!j~ZS Maintaining an oparat'eakage limit of 150 gpd per steam ganeratoz (600 gpd total) for Fuel Cycle vill minimize the potential for a large leakage event during steam line break under LOCA conditions.

Based on the NDE uncertainties, bobbin coil voltage distzibution and czack growth rata from the previous inspection, the expected leak rate following a steam linn rupture is Limited to helov 12.6 gpm vhich willlimit the calculated offsite doses to vithin 10 percent of 10 CFR 100 guidelines.

Leakage in the intac loops is limited to 150 gpd.

Zf the projered end of cycle distribution of crack indications results in primary-to-secondary leakage gzaatar than 12.6 gpm in the faulted loop during postulated steam line break event, additional tubes must he removed from service in order to reduce the postulated primary-to-secondary steam line break leakage to below 12.6 gpm.

PRESSURE BOUNDARY LZAKACZof any magnitude is unacceptahla since it may be indicative of an impending gross failure of tha pressure boundary.

Should

.PRESSURE BOUNDARY LEAKAGE occur through a component which can he isolated f om the balance of the Reactor Coolant System, plant operation may continue provided the leak'ng component is promptly isolated from the Reactoz Coolant System since isclat'on removes the souzce of potential failure.

3 4. 4.7 CH:-MTSTRY The limitations on Reac=or Coolant System chemistry ensure that corrosion o

the Reactor Coolant System is minimized and reduces the potential for Raac=or Coo'an" System leakage or failure dua to stress corrosion.

Maintaining the chemistg within the Steady State Limits provides adequate coz"osion protect'on to ensure the st= crural integrity of dna Reactor Coolant System over the life o

tha plant.

The associated effare of exceeding the oxygen,

chlozida, and fluor'de limits are time and temperature dependa'nt.

Corrosion studies show that opezat'on may be continued with contaminant concentration levels in excess of the Steady State Limi s, up to the Transient Limits, foz the specified limited time intervals vithout having a significant effect on the structural intagz'ty of the Reac=cr Coolant System.

The time intazval permitting continued operation within the rastz'c iona of the Transient Limits provides time for taking corzective ac=ions to restore the contaminant concentrations to within ~ the Steady State Z.imi s.

COOK NUCLEAR PLANT UNIT l B 3/4 4-4 AY NDHZNT NO. ~, 444, ~

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~co~<<

8v~ <o 9(<<9 3/4 0-f The su~iI.Lance ac@xi-events ozovide adequate assu=ance ut concen"=scions excess oz the Li 'its vtLL be detected ln su"=icient t e to take corrective action.

<.4.8 S vC The Limitations on the specific activfty of the p&~ coolant ensu=e t."~t the result~kg 2 hou doses at the sita bounds~ vtLL not axceed an appropriately snaLL f~ction of Pa,L00 L~ts foLLo&mg a stean ge ezatoz rube rupture accidc fm con]unction vith an assuned steady state pr~~~

to-secondary stean generator Leakage rate of 1.0 CPS.

The values for the L~iits on specific activi~ represent interim LMts based upon a para=auric evaluation by'he NRC of typicaI. site locations.

These values are conservative in that specific site pazaneters of the Cook Huclear Plant sf.te, such as site bounda~ location and meteorological cond't'ons, vere

"..ot considered in this evaluation.

The HRC is f~~lf=img sita spec~='c c-'----'a.

vhich vtLL be used as the basis for the reevaluation of the spec'=ic activity Lmiits of this site. This reevaluation nay resuLt M h'ghe 1

'--'89site

'doses follovtng a nai< stean Line bzeak aze Lwi'ted to LO perte"t o"

'he 10 ~c 100 guidelm~e.

The restriction is based on a Cook Rxclear Plant sita-spec~~ic radiological evaluation that as~as a post-accident pr~~-

to-seconda~

leak zata of 120 gpss iu the fauLted Loop and a pz~~ coolant specific activity concen~tion corresponding zo LX fuel defects (appro~teLy 4.6 nic=oCu=dies/~

dose e~valent Z-L31), rathe than a spec~~ic active.g of 1.0 niczoCuzies"dose equivalent X-131.

Reduce~

g T~ to less zhan 500 F pzevents the release of activizy shouLd a

o s ean gene ator tube zuptu e since the satu ation prassu"e of the pz~~=z coolant is belov the Lift pressu"e of ohe atmospheric stean relief valves The suweil'ance za~enents provide adecpzata assu=ance that excessive spec ~ic activi~ leveLs in the pri~~ coolant vtLL be detected M

su -icient ti~e to oi3ce cor=ective ac ion.

In="orat'on obta~

ed on iod='e spQ-ag vill be used to assess zhe pa<<sneters associated vith summing phenonena.

h reduction in frequency of isotopic anaLyses foLLovimg pover changes nay be pe~issible i" just~"ied by the data obtained.

CQQK ÃJC~R PLAVZ - VNXT 1 B 3/4 4-5 h~~

HO. ~

Loo~

ATTACHMENT 3 TO AEP:NRC:1166R PROPOSED REVISED TECHNICAL SPECIFICATION PAGES

I

3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS

~

3/4 4'EACTOR COOLANT SYSTEM STEAM GENERATORS LIMITINGCONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3 and 4.

ACTION:

With one or more steam generators inoperable, restore the inoperable generator(s) to OPERABLE status prior to increasing T>>< above 200'F.

SURVEILLANCERE UIREMENTS 4.4.5.0 1

Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirement of Specification 4.0.5.

4.4.5.1 Steam Generator Sam le Selection and Ins ection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.

4.4.5.2 Steam Generator Tube Sam le Selection and Ins ection - The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2.

The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4.

The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:

Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% ofthe tubes inspected shall be from these critical areas.

The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:

1.

Alltubes that previously had detectable wall penetrations (greater than or equal to 20%) that have not been plugged or repaired by sleeving in the affected area.

This Specification does not apply in Mode 4 while performing crevice fiushing as long as Limiting Conditions for Operation for Specification 3.4.1.3 are maintained.

COOK NUCLEAR PLAiT-UNIT1 Page 3/4 4-7 AMENDMENT4Q, 446

3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCERE UIREMENTS (continued) 2.

Tubes in those areas where experience has indicated potential problems.

A tube inspection (pursuant to Specification 4.4.5.4.a.8) shall be performed on each selected tube. Ifany selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.

4.

Tubes left in service as a result of application of the tube support plate plugging criteria shall be inspected by bobbin coil probe during all future refueling outages.

c.

In addition to the sample required in 4.4.5.2.b.1 through 3, all tubes which have had the F>> criteria applied will be inspected in the roll expanded region.

The roll expanded region 'of these tubes may be excluded from the requirements of 4.4.5.2.b.l.

d.

The tubes selected as the second and third samples (ifrequired by Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:

1.

The tubes selected for the samples include the tubes from those areas ofthe tube sheet array where tubes with imperfections were previously found.

2.

The inspections include those portions of the tubes where imperfections were previously found.

Implementation of the steam generator tube/tube support plate plugging criteria for one fuel cycle (Cycle 15) requires a 100 percent bobbin coil inspection for hot-leg tube support plate intersections and cold-leg intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications.

The determination of tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20 percent random sampling of tubes inspected over their full length.

The results of each sample inspection shall be classified into one of the following three categories:

Cater Ins ection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.

C-3 More than 10% of the total tubes inspected are degraded tubes nr more than 1%

of the inspected tubes are defective.

COOK NUCLEAR PLANT-UNIT1 Page 3/4 4-8 AMENDMENT4$4, 446, ~) K8

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3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS a

3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCERE UIREMENTS (continued)

Note:

In all inspections, previously degraded tubes must exhibit significant (greater than or equal to 10%) further wall penetrations to be included in the above percentage calculations.

4.4.5.3 Ins ection Fre uencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:

The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality. Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections followingservice under AVTconditions, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months.

b.

If the results of inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40 month intervals fall in Categoty C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3.a; the interval may then be extended to a maximum of once per 40 months.

Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:

1.

Primary-to-secondary tubes leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2.

2.

A seismic occurrence greater than the Operating Basis Earthquake.

3.

A loss-of-coolant accident requiring actuation of the engineered safeguards.

4.

A main steam line or feedwater line break.

COOK NUCLEAR PLANT-UNIT1 Page 3/4 4-9 Ai~NDMENT98, 446

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sag 3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.4'EACTOR COOLANT SYSTEM SURVEILLANCERE UIREMENTS (continued) 4.4.5.4 Acce tance Criteria As used in this Specification:

~tm erfection means an exception to the dimensions, finish or contour of a tube or sleeve from that required by fabrication drawings or specifications.

Eddy~rrent testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.

~De radation means a service-induced

cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube or sleeve.

De raded Tube or Sleeve means an imperfection greater than or equal to 20% of the nominal wall thickness caused by degradation.

Percent De radation means the amount of the tube wall thickness affected or removed by degradation.

Defect means an imperfection of such severity that it exceeds the repair limit.

6.

Re air/Plu in Limit means the imperfection depth at or beyond which the tube or sleeved tube shall be repaired or removed from service.

Any tube which, upon inspection, exhibits tube wall degradation of40 percent or more ofthe nominal tube wall thickness shall be plugged or repaired prior to returning the steam generator to service.

This definition does not apply to the portion of the tube in the tubesheet below the Fd'istance for F* tubes.

Any sleeve which, upon inspection, exhibits wall degradation of 29 percent or more of the nominal wall thickness shall be plugged prior to returning the steam generator to service.

In addition, any sleeve exhibiting any measurable wall loss in sleeve expansion transition or weld zones shall be plugged.

This definition does not apply to tube support plate intersections for which the voltage-based plugging criteria are being applied.

Refer to 4.4.5.4.a.l0 for the plugging limit applicable to these intersections.

Unserviceable describes the condition of a tube or sleeve ifit leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3.c, above.

~tns ection determines the condition of the steam generator tube or sleeve from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.

For a tube in which the tube support plate elevation interim plugging limit has been

applied, the inspection will include all the hot leg intersections and all cold leg intersections down to, at least, the level of the last crack indication.

g~leevin a tube is permined only in areas where Ihe sleeve spans the tubesheet area and whose lower joint is at the primary fluid tubesheet face.

COOK NUCLEAR PLANT-UNIT1 Page 3/4 4-10 AMENDMENT08, 4R, 446) ~

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3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.4 ~

REACTOR COOLANT SYSTEM SURVEILLANCERE UIREMENTS (continued) 10.

Tube Su ort Plate Re air Limit is used for the disposition of a steam generator tube for continued service that is experiencing outside diameter stress corrosion cracking (ODSCC) confined within the thickness of the tube support plates. At tube support plate intersections, the repair limit is based on maintaining steam generator tube serviceability as described below:

a.

Degradation attributed to ODSCC within the bounds of the tube support plate with bobbin voltage less than or equal to 2.0 volts willbe allowed to remain in service.

b.

Degradation attributed to ODSCC within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts willbe repaired or plugged except as noted in 4.4.5.4.a. 10.c below.

~ t Indications of potential degradation attributed to ODSCC within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts but less than or equal to 5.6 volts may remain in service ifa rotating pancake coil inspection does not detect degradation.

Indications of ODSCC degradation with a bobbin voltage greater than 5.6 volts willbe plugged or repaired.

11.

F* Distance is the distance from the bottom of the hardroll transition toward the bottom of the tubesheet that has been conservatively determined to be 1.11 inches (not including eddy current uncertainty).

12.

F~ Tube is a tube with degradation, below the F~ distance, equal to or greater than 40%,

and not degraded (i.e., no indications of cracking) within the F~ distance.

The steam generator shall be determined OPERABLE after completing the corresponding actions (plugging or sleeving all tubes exceeding the repair limit and all tubes containing through-wall cracks) required by Table 4.4-2.

Steam generator tube repairs may be made in accordance with the methods described in either WCAP-12623 or CEN-313-P.

COOK NUCLEAR PLANT-UNIT1 Page 3/4 4-11 AMENDMENT4', 466, ~, XV'

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3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS

c. 3/4 4 REACTOR COOLANT SYSTEM SURVEILLANCERE UIREMENTS (continued) 4.4.5.5 hearts Following each inservice inspection of steam generator tubes, if there are any tubes requiring plugging or sleeving, the number of tubes plugged or sleeved in each steam generator shall be reported to the Commission within 15 days.

The complete results of the steam generator tube inservice inspection shall be included in the Annual Operating Report for the period in which this inspection was completed.

This report shall include:

1.

Number and extent of tubes inspected.

2.

Location and percent ofwall-thickness penetration for each indication ofan imperfection.

3.

Identification of tubes plugged or sleeved.

C.

Results of steam generator tube inspections which fall into Category C-3 and require prompt notification of the Commission shall be reported pursuant to Specification 6.9.1 prior to resumption of plant operation.

The written followup of this report shall provide a description of investigations conducted to determine cause ofthe tube degradation and corrective measures taken to prevent recurrence.

For implementation of the voltage-based repair criteria to tube support plate intersections, notify the staff prior to returning the steam generators to service should any of the following conditions arise:

Ifestimated leakage based on the actual measured endwf~cle voltage distribution would have exceeded the leak limit(for the postulated main steam line break utilizing "Standard Review Plan - NUREG-0800" assumptions) during the previous operating cycle.

If circumferential crack-like indications are detected at the tube support plate intersections.

Ifsignificant indications are identified that extend beyond the confines ofthe tube support plate.

Ifthe calculated conditional burst probability, as calculated per WCAP-14277, exceeds Ix 10z, notify the NRC and provide an assessment of the safety significance of the occurrence.

COOK NUCLEAR PLANT-UNIT1 Page 3/4 4-12 AMENDMENT4Q, 446, ~, 478

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8 3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.4" REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITINGCONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a.

No PRESSURE BOUNDARYLEAKAGE, b.

1 GPM UNIDENTIFIEDLEAKAGE, c.

600 gallons per day total primary-to-secondary leakage through all steam generators and 150

-'allons per day through any one steam generator for Fuel Cycle 15, d.

10 GPM IDENTIFIEDLEAKAGEfrom the Reactor Coolant System, m

e.

Seal line resistance greater than or equal to 2.27E-I ft/gpm~ and, f.

The leakage from each Reactor Coolant System Pressure Isolation Valves specified in Table 3.4-0 shall be limited to 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm, at a Reactor Coolant System average pressure within 20 psi of the nominal full pressure value.

APPLICABILITY: MODES 1, 2, 3 Bnd 4.

ACTION:

With any PRESSURE BOUNDARYLEAKAGE,be in at least HOT STANDBYwithin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARYLEAKAGE,reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

With any reactor coolant system pressure isolation valve(s) leakage greater than the above limit, declare the leaking valve inoperable and isolate the high pressure portion of the affected system from the low pressure portion by the use of a combination of at least two closed valves, one of which may be the OPERABLE check valve and the other a closed de-energized motor operated valve. Verify the isolated condition of the closed de-energized motor operated valve at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Specification 3.4.6.2.e is applicable with average pressure within20 psi ofthe nominal fullpressure value.

COOK NUCLEAR PLANT-UNIT1 Page 3/4 4-16 AMENDMENT443) 446) 474) 488

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3/4 BASES s 3/4.4

~ REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS TUBE INTEGRITY The Surveillance Requirements for inspection ofthe steam generator tubes ensure that the structural integrity of this portion of the RCS willbe maintained.

The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.

Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is expected to be operated in a manner such that the secondary coolant willbe maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.

Ifthe secondary coolant chemistry is not maintained within these parameter limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation ofsteam generator tube leakage between the primary coolant system and the secondary coolant system. The allowable primary-to-secondary leak rate is 150 gallons per day per steam generator for one fuel cycle (Cycle 15).

Axial or circumferentially oriented cracks having a primary-to-secondary leakage less than this limit during operation willhave an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents.

Leakage in excess of this limit willrequire plant shutdown and an inspection, during which the leaking tubes willbe located and plugged or repaired.

A steam generator while undergoing crevice flushing in Mode 4 is available for decay heat removal and is operable/ operating upon reinstatement ofauxiliary or main feed flowcontrol and steam control.

Wastage-type defects are unlikely with the all volatile treatment (AVT)of secondary coolant.

However, even if a defect of similar type should develop in service, it willbe found during scheduled inservice steain generator tube examinations.

Plugging or sleeving will be required for all tubes with imperfections exceeding the repair limit which is defined in Specification 4.4.5.4.a.

Steam generator tube inspections ofoperating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.

Tubes experiencing outer diameter stress corrosion cracking within the thickness of the tube support plates are plugged or repaired by the criteria of 4.4.5.4.a. 10.

COOK iNUCLEARPLANT-UNIT1 Page B 3/4 4-2a AMENDMENT4Q, 484, 446) 478

3/4 BASES c.'/4,4>>

REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS TUBE INTEGRITY Continued Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Commission pursuant to Specification 6.9.1 prior to resumption ofplant operation. Such cases willbe considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

Degraded steam generator tubes may be repaired by the installation of sleeves which span the section of degraded steam generator tubing. A steam generator tube with a sleeve installed meets the structural requirements of tubes which are not degraded.

To determine the basis for the sleeve plugging limit, the minimum sleeve wall thickness was calculated in accordance with Draft Regulatory Guide 1.121 (August 1976).

In addition, a combined allowance of 20 percent of wall thickness is assumed for eddy current testing inaccuracies and continued operational degradation per Draft Regulatory Guide 1.121 (August 1976).

The following sleeve designs have been found acceptable by the NRC staff:

1.

Westinghouse Mechanical Sleeves (WCAP-12623) 2.

Combustion Engineering Leak Tight Sleeves (CEN-313-P)

Descriptions of other future sleeve designs shall be submitted to the NRC for review and approval in accordance with 10 CFR 50.90 prior to their use in the repair of degraded steam generator tubes.

The submittals related to other sleeve designs shall be made at least 90 days prior to use.

COOK NUCLEAR PLANT-UNITI Page B 3/4 4-2b AMENDjlIENT4N, 484 446

3/4 BASES 3/4 4~

REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGEDETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary.

These detection systems are consistent with the recommendations of Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.

3/4.4.6.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than I gpm. This threshold value is sufficiently low to ensure early detection of additional leakage.

The 10 GPM IDENTIFIEDLEAKAGElimitations provides allowance for a limited amount of leakage from know sources whose presence will not.interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.

The limitation on seal line resistance ensures that the seal line resistance is greater than or equal to the resistance assumed in the minimum safeguards LOCA analysis.

This analysis assumes that all of the flow that is diverted from the boron injection line to the seal injection line is unavailable for core cooling.

Maintaining an operating leakage limit of 150 gpd per steam generator (600 gpd total) for Fuel Cycle 15 will minimize the potential for a large leakage event during steam line break under LOCA conditions.

Based on the NDEuncertainties, bobbin coil voltage distribution and crack growth rate from the previous inspection, the expected leak rate following a steam line rupture is limited to below 12.6 gpm which willlimitthe calculated offsite doses to within 10 percent of the 10 CFR 100 guidelines.

Leakage in the intact loops is limited to 150 gpd. Ifthe projected end of cycle distribution of crack indications results in primary-to-secondary leakage greater than 12.6 gpm in the faulted loop during a postulated steam line break event, additional tubes must be removed from service in order to reduce the postulated primary-to-secondary steam line break leakage to below 12.6 gpm.

PRESSURE BOUNDARYLEAKAGEof any magnitude is unacceptable since it may be indicative ofan impending gross failure ofthe pressure boundaty.

Should PRESSURE BOUNDARYLEAKAGEoccur through a component which can be isolated from the balance of the Reactor Coolant System, plant operation may continue provided the leaking component is promptly isolated from the Reactor Coolant System since isolation removes the source of potential failure.

COOK NUCLEAR PLANT-UNIT1 Page B 3/4 4-3 AMENDivtENT&, 6k, 4A) 4A Corrected Page

3/4 BASES X4,4i REACTOR COOLANT SYSTEM 3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion.

Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant.

The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent.

Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System.

The time interval permitting continued operation within the restrictions of the Transient Limitsprovides time fortaking corrective actions to restore the contaminant concentrations to withinthe Steady State Limits.

The surveillance requirements provide adequate assurance that concentrations in excess ofthe limits willbe detected in sufficient time to take corrective action.

COOK NUCLEAR PLANT-UNITI Page B 3/4 44 AMENDMEVI'4k,446, 478, 488

3/4 BASES

~', $/4 4.

REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary willnot exceed an appropriately small fraction ofPart 100 limits followinga steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM.

The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations.

These values are conservative in that specific site parameters of the Cook Nuclear Plant site, such as site boundary location and meteorological conditions, were not considered in this evaluation. The NRC is finalizing site specific criteria which willbe used as the basis for the reevaluation of the specific activity limits of this site. This reevaluation may result in higher limits.

Offsite doses following a-main steam line break are limited to 10 percent of the 10 CFR 100 guideline.

The restriction is based on a Cook Nuclear Plant site-specific radiological evaluation that assumes a post-accident prirnaiy-to-secondary leak rate of 120 gpm in the faulted loop and a primary coolant specific activity concentration corresponding to 1% fuel defects (approximately 4.6 microCuries/gram dose equivalent 1-131), rather than a specific activity of 1.0 microCuries dose equivalent I-131.

Reducing T>>< to less than 500'F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the liftpressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant willbe detected in sufficient time to take corrective action. Information obtained on iodine spiking willbe used to assess the parameters associated with spiking phenomena.

A reduction in frequency of isotopic analyses following power clianges may be permissible ifjustified by the data obtained.

COOK NUCLEAR PLANT-UNIT1 Page B 3/4 4-5 AMENDMENT44k, 446

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