ML20133M554

Proposed Tech Specs Implementing performance-based Containment Leak Rate Testing Requirements of 10CFR50, App J,Option B

ML20133M554
Person / Time
Site:	McGuire, Mcguire
Issue date:	01/13/1997
From:	DUKE POWER CO.
To:
Shared Package
ML20133M550	List: ML20133M549 ML20133M554
References
NUDOCS 9701220425
Download: ML20133M554 (35)

Text

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Attachment I Technical Specification Mark-ups and Revised Originals 1

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j 9701220425 970113 '

PDR ADOCK 05000369 P PDR

_- - - - - - ~ ~ ~ ' ~ ~ ~ ^ ~ '

e 3/4.6 CONTAINMENT SYSTEMS i ) E Wls PA6Ej

! ' 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY

' ,i 9N OMQ LIMITING CONDITION FOR OPERATION 3.6.1.1 '\

Primary CONTAINMENT INTEGRITY shall be maintainted.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

'Without primary CONTAINMENT INTEGRITY, restore CONTAINMEN SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.I hour or be inr.at least HO

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SURVEILLANCE REQUIREMENTS 4.6.1.1 i Primary CONTAINHENT INTEGRITY shall be demonstrated:

a.

At least once per 31 days by verifying that all penetrations

• not .

.) valves or operator action during periods when co valves are open under administrative control,** and required to be i closed during accident conditions are closed by valves, blind  !

flanges, or deactivated automatic valves secured in their position!

b. t Specification. 3.6.1.3; andBy verifying that each containment air lo l c.

After each closing of each penetration subject to Type B testing except the containment air locks, if opened following a Type A or B test, by leak rate testing the seal with gas at P added to the leakage rates determined pursuant to Sve t '

4.6.1.2d. for all other Type B and C penetrations, the pecification combined leakage rate is less than 0.60 L,.

• Except valves, blind flanges, and deactivated autom otherwise secured in the closed position.

l be performed more often than once per 92 days.fied close T

/ **The trative following control:valves may be opened on an intermittent basis under adminis-FW-4. NC-141, NC-142, WE-13, WE-23, VX-34, VX-40, FW-11, FW-13, McGUIRE - UNIT 1 3/4 6-1 Amendment No. 166

CONTAINHENT SYSTEMS

) CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:

a.

An overall integrated leakage rate ofg g c

p gess than or equal to L,, 0.30% by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at Pa ,14.8 psig, se

--2 ) Lass than-or-equaLto I g 0.14% by weight of the containment

-a4r-- per 2d. houn, at a ri:duced pi mssure vi Pt

• 7 PSit-

b. A combined leakage rate of less than 0.60 L for all penetrations a

and and valves subject to Type B and C tests, when pressurized to P,,

c.

A combined bypass leakage rate of less than 0.07 L, for all penetra-tions identified pressurized to P as- secondary containment bypass leakage paths when a

APPLICABILITY: MODES 1, 2, 3, and 4.

) ACTION:

With 0.75 L(a) the measured overall integrated containment leakage rate exceeding M'E ' gr app 14eaM% or (b) the measured combined leakage rate for all penetrations and valves subject to Types B and C tests L,, exceeding 0 or (c) the combined bypass leakage rate exceeding 0.07 L restore the -

overall integrated leakage rate to less than 0.7517 3oF-tess,,han t .or--ettual--ti f 75 L ,t as applicaMg and the combined leakage rate for all penetrations and valvesleakage subject , and the combined L,i bypass rate to Type to less than80.07andLa C tests Prior to increas to less ng thethan 0.60 Reactor Coolant System temperature above 200'F SURVEILLANCE REQUIREMENTS 4.6.1.2 W A OcWBIL}MFJJT hracrit le6kege-rates LEAKAf6 chall k "Park $ sHna 13s

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M;.>. 4 ar2-4L the iisa55-fr}9% go$,hgfg' n "m.a~u u:> ing we--methods-and-provis4cne hEff6AJ9fMf6b AS 86&U//E& hY /O 0fkNDo$'th) Y AblUIblY oc to CFR.50,OPf/bA)6 3 M MbiFl66 0f' /WPROV6 6)<6/fN/O f%Jd )A) ACYWY UlY Tk QV/.b6L//lbb b

of tecsvurrotu' Gotbc Q63)JcPf9106Q Ipjf McGUIRE - UNIT 1 3/4 6-2 Amendment do. 166 l

CONTAINMENT SYSTEMS

) SURVEILLANCE REQUIREMENTS (Continued) a./ sh 11 be cree Ty) A test (OveraQIntegray Containment Lhkage Rate)

BQ.gf@ eit r P,, 1ducted t 40 i I month in rvals 'd ring shqtdown at 8 psig, r at P t , 7.4 psig, during ach 10 'y ar N servi perio x The th rd test f each se shall cond d during he shutMown for the 10-yeag plant in ervice 'nspecti  ;

b. any perkdic Type A est fails

• apprN t 1{hhtest ,

scheb lee for su meet eithr 0.75 L, r 0.75 L

• t o d by the ommission.quent Type tests sh 11 be rev' wed and meet ei If two cons cutive Ty A tests ail to r 0.75 L, or 0.75 L t , Type A t shall b erforme at st ever 18 0.7 L, or 0. 5 L at month until two secutive pe A test meeteither resum ; t ich time th above tes chedule be c.

The accuracy of each Type A test shall be verified by a supplemental test

1) - Cwhnh+ IM fl%M OIT)/ %duYrctw' GUthe f.ts3]MBCA s d h h acenrac of the Tyne A test-Ay verifying tha@ 'M8 diff ence b tween upplemB tal a 0.25 L, or 0. 5 Lt i Type A est data 's within 3

2) as a du L{ekageration l betws nfficie the t to est lish a uratel the chang in pe A te t and t and, supple ental tes

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) 3) Requi es the gyantit of gas 'njected 'nto the ntainme or bled f' the chntainm nt duri g the su lemental test to

%8equivalen psid, urtoathleast P t* 2 4 W- 55; of th) total , asured 1 kage at

d. Type B and C tests shall be conducte (Wg'as wit M Ud4.8 P,

Wsg/ Io#N8)

W intervals no greater than 24 months.except for tests involving: A;

1) Air ' locks,

2) .

Dual-ply bellows assemblies on containment penetrations between the containment building and the annulus, and

3) Purge supply and exhaust isolation valves with resilient material seals.

4)

Type C tests performed on containment penetrations M372, M373 without draining the glycol-water mixtur from the seats of their diaphragm valves (NF-228A, NF-233B, and NF-234A), if meeting a zero indicated leakage rate (not including instrument error) for the diaphragm valves. These tests may be used in lieu of tests which are otherwise required by Section III.C.2(a) of 10 CFR 50, Appendix J to use air or nitrogen as McGUIRE - UNIT 1 3/4 6-3 Amendment No. 166 i

_ --- - - ~ -- - ~ ' ' ~ ~~~~~~

CONTAINMENT SYSTEMS i ')

SURVEILLANCE REQUIREMENTS (Continued) the test medium. The above required test pressure (P3 ) and test interval are not changed by this exception.

e.

Purge supply and exhaust isolation valves with resilient material I seals shall be tested and demonstrated OPERABLE by the requirements of Specification 4.6.1.9.3 or 4.6.1.9.4, as applicable; f.

The combined bypass leakage rate shall be determined ,to be less than '

O.07 except L,for penetrations which are not i di idby applicable n v ually testable; Type penetrations not individually testable shall be detennined to have no detectable leakage when tested with soap bubbles while the con-tainment is pressurized to P ,14.8 psigg og ' 4 ps49, during each Type A test; .

l g.

Air locks shall be tested and demonstrated OPERABLE per Specification 4.6.1.3;

h. i The space between each dual-ply bellows assembly on containment penetrations between the containment building and the annulus ~shall be vented to the annulus during Type A tests. Following completion of each Type A test, the space between each dual-ply bellows

) assembly shall be subjected to a low pressure test at 3-5 psig to verify no detectable leakage or the dual-ply bellows asserlaly shall be subjected to a leak test with the pressure on the containment side of the dual-ply bellows assembly at P ps49 , to verify the leakage to be within t$e,14.8 limits psig, c P , 7.4 of Specification'

4. 6.1.2f. ;

i.

All test leakage rates shall be calculated using observed data converted to absolute values. Error analyses shall be performed to select a balanced Integrated Leakage Measurement System; and j.

The provisions of Specification 4.0.2 are not applicable.

1 i

McGUIRE - UNIT 1 3/4 6-4 Amendment No. 166

_w,.ns.s-' E->#'* ^ ' ^ ._ _ _ ___ _- ---- ~~

CONTAINMENT SYSTEMS

) CONTAINMENT VESSEL STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.6 The structural integrity of the containment vessel shall be maintained 4.6.1.6. at a level consistent with the acceptance criteria in Specification APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With the structural integrity of the containment vessel not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200*F.

SURVEILLANCE REQUIREMENTS 4.6.1.6 The structural integrity of the containment vessel shall be determined-dur4ncy-tb shutdown fer ecch Type ? containment-lechge-rate-test (rcfcrence SpecMissuen 4.5.1 A!4 by a visual inspection of the exposed accessible interior and exterior surfaces of the vessel. This inspection

-) shall be performed prior to the Type A containment leakage rate testMo verify no apparent changes in appearance of the surfaces or other abnormal degradation.# Any abnormal degradation of the containment vessel detected t

during the above required inspections shall be reported to the Commission pursuant to 10 CFR Sections 50.72 and 50.73.

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, McGUIRE - UNIT 1 3/4 6-9 Amendment No. 166 l

CONTAINMENT SYSTEMS _

i REACTOR BUILDING STRUCTURAL INTEGRITY

. LIMITING CONDITION FOR OPERATION 3.6.1.7 The structural integrity of the reactor building shall be maintai at a level consistent with the acceptance criteria in.... Specification ned 4 61 APPLICABILITY: H0 DES 1, 2, 3, and 4.

ACTION:

With the structural. integrity of the reactor building not conf orming to the prior to increasing the Reactor Coolant m s System tem

_ SURVEILLANCE REQUIREMENTS 4.6.1.7 during the shutdown for each Type A containment Specification 4.6.1.2) by a visual inspection of accessible the exposed ermined reference le interior and exterior surfaces of the reactor bailding verifying and no or other ab apparent changes in appearance of the concrete surfaces radation normal i

l e a ove required inspections shall be ursuant 10 CFR Sections 50.72, and 50.73. reported e uring to to the C

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3/4.6 CONTAINMENT SYSTEMS

)

BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive paths and associated leak rates assumed in the This accident analyses restriction, in conjunction with the leakage rate limitation, will limit the site boundary Part 100 during radiation accidentdoses to within the dose guideline values of 10 CFR conditions. .

3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure, P . As an added conservatism, the measured equal to 0.75 overall L '"integrated

• * ' leakage rate ,is further limited to less than or a

teststoaccountforpossib5 "5 "I,p koirk,. during perfonnance of the periodic between leakage tests. le degradation of the containment leakage barriers

,f'TS'PC A ,S The surveillance testing for measuring leakage rates ere consistent with the requirements of Appendix J of 10 CFR 50, op7 fog 3, 7ype dg.st5 ,

gug C 3 .6. 3 0 NME T AI LO S O FO MM/M Gj Ng / A, The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate.

Surveillance testing of the air lock seals provide assurance that the overall air lock the intervals between leakage will-not air lock become leakage tests.excessive due to seal ~ damage during l

McGUIRE - UNIT 1 B 3/4 6-1

- _ - _ _ - _ - - _ - - - - ~

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_ CONTAINMENT SYSTEMS BASES 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that:

(1) the differential with respect to the outside, atmosphere and (2) the of 1.5 containment LOCA conditions. peak pressure does not exceed the design pressure of 15 i The 14.5 psig. maximum peak pressure expected to be obtained from a LOCA event The limit of 0.3 psig for initial positive containment pressure will limit the total pressure to 14.8 psig which is less than the design pressure and is consistent with the accident analyses.

3/4.6.1.5 AIR TEMPERATURE (1) the containment air mass is limited to an initial mass su ambient air temperature does not exceedthethat temperatu continuous within duty containment. rating specified for equipment and instrumentation located Measurements shall be made at all listed locations whether temperature.by fixed or portable instruments, prior to determining the avera, air

)

air mass during a LOCA.The containment pressure transient ned is sensitive to th perature.

The lower temperature limit of 100*F for the lower compa for the upper compartment and 60*F when less than or equal , 75*F to 5% of RATE THERMAL POWER will limit the peak pressure to 14.8 psig n the which is less tha

. the peak accident -temperature 'siightly'duri'ncj a LOC Both used in thethe upper accident analyses.and lower temperaturemeters limits ons. are con 3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containme t steel the lifevessel will be maintained comparable to the originaln design standard of the facility.

vessel will withstand the maximum pressure of 15 psig in the eve Aemonstrate visual inspection -in-conjenetien ith Type ^ leigotcsti is suffici this capability. ent to PcAloblG

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3/4.6 CONTAINMENT SYSTEMS No CHAUGES THLS PMyk

) 3/4.6.1 PRIMARY CONTAINMENT jgMDATW ONL// v c

CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION

3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

I APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

l ,

4 Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within j 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be SHUTDOWN in at wif.hin theleast HOT following STANDBY 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. 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 O

i SURVEILLANCE REQUIRFMENTS h/

Ii h;

g l 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated: V

a. [Q At least once per 31 days by verifying that all penetrations

• not Si capable of being closed by OPERABLE containment automatic isolation sti

) valves or operator action during periods when containment isolation M valves are open under administrative control,** and required to be W closed during accident conditions are closed by valves, blind W flanges, or deactivated automatic valves secured in their positions; k.$

b.

By verifying that Specification each and 3.6.1.3; containment air lock is in compliance with c.

After each closing of each penetration subject to Type B testing, except the containment air locks, if opened following a Type A or B test, by leak rate testing the seal with gas at P 14.8 psig, and verifying that when the measured leakage rate for,,hese t seals is

! added to the leakage rates determined pursuant to S i 4.61.2d. for all other Type B and C penetrations, the pecification combined t leakage rate is less than 0.60 L,.

)

i *Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and the annulus and are locked, sealed or l

otherwise secured in the closed position. These penetrations shall be veri-fied closed during each COLD SHUTDOWN except that such verification need not be perfonned more often than once per 92 days.

l T ,

j **The following trative valves may be opened on an intermittent basis under adminis-control: @'c FW-4. NC-141, NC-142, WE-13, WE-23, VX-34, VX-40, FW-11, FW-13, p!!

) ty

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McGUIRE - UNIT 2 3/4 6-1 Amendment No. 148 d

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s i CONTAINMENT SYSTEMS !G.,

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) CONTAINMENT LEAKAGE ,

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_ LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:

I L

a. An overall integrated leakage rate ofs

,0 weight i

p ~M7 air hessper 24than hours or equal to , Lep-at P,, .

14 8of the psigcontainment j =2)

Less than or ennal to I g n la by weight-bf u,e contoin;acnt-g l' nr per 2? hours at e reduced pressure sfv7.4 ps? ir.

6 b. A combined leakage rate of less than 0.60 L for all penetrations t

and valves subject to Type B and C tests, w$en pressurized to P ,

and c.

4 A combined bypass leakage rate of less than 0.07 L, for all penetra-

! tions identified pressurized to as secondary containment bypass leakage paths when P,.

4 1

APPLICABILITY: MODES 1, 2, 3, and 4.

a

) ACTION:

j

] With (a) the measured overall integrated containment leakage rate exceeding

.;i 0.75 L,, L on-0J54,-as-appMcaMe7 or (b) the measured combine e j or (c) the combined bypass leakage rate exceeding 0.07 L overall integrated leakage rate to less than 0375-L a oMeM,, restore the an-or-epaW ~

)q -A254, as ap'pMeabh, and the combined leakage rate for all penetrations and i valves subject to Type B and C tests to less than 0.60 L , and the combined bypass leakage rate to less than 0.07 L, prior to increa, sing the Reactor Coolant System temperature above 200*F l

1 SURVEILLANCE REQUIREMENTS i

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4 4.6.1.2 Ge containment leakage rates shall be demonstrated et the ,-xxuurAs8 Q R6dO h schedM a and sha11 be dete -ined in l fied in Appendia J JMCF", 50 using 4he methnd previsicas-#Att31coafer-aa.ce itLt 4

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M CONTAINHENT SYSTEMS '

A r ) SURVEILLANCE REQUIREMENTS (Continued) a.

Three Type A test (OvePa 1 Integ ated Conta s(allbeloqducted 40 1 ent Leab ge Rate) t A ei ttt r P,,14 8 psig, mont ' tervals d 'ng shutM n at b6L v servi period. The thi at Pt, test o

.4 ps during ea 10-during t e shutdo for th 10-year ch se shall be con (ye i ucte nt in vice inspebt. ion;

) b. If qny perl p thetstsc%o'cTypeAt t fails meet eith 0.75 L, 0.75 L*

i approv heda e for subs by the ission. yent Type tests sha be revi t i 'I ad and eet eit r 0.75 L, r 0.75 Lt . two cons utive Typ A tests il to ,

W

(,

le t ever Type A te t shall be erformed t f

[ 0.7 or 0. 8 months ntil two c secutive L at w 'ch time th abovetestychedulem t

ype A test meet eit r Q resume ; be R

( c. R The accuracy of each Type A test shall be verified by a supplemental -

whiM1 4 Acce>({fjy/CG 4)tTH /2G6UUW test.Confi O RY(Sulb6 I,/(o3 S d

h the a uracy. of -t 3 differen e betwe n supplemen -Type A tes by verifying hat the 19 7.5 w '

al and Type 0.25 L,, 0.25 - test data is ithin R:

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% s a durati y y le kage rate b tweensuffi ient to est (lish accurat ly the chang in 4

n y and e Type A tes and the su lemental tes ;

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% tg 3 Requi s the quant y of as injected bled fr the contai ment to the conta' ment or l equivale $ to at leas 25% ring the su lemental tes to be yi-i 14.8 psig,j or P , 7.4 p ig. { the total asured leakag at P,, h>l d.

t Type B and C tests shall be conducted g

gas .at P,, 14.8 psig, at -

3 intervals no greater .than 24 monthsc ex' ept for tests involving:

1) Air locks, 2)

Dual-ply bellows assemblies on containment penetrations between the containment building and the annulus, and +

3)

Purge supply and exhaust isolation valves with resilient l material seals. v.

4) 1 b I

Type C tests performed on containment penetrations M372, M373 without draining the glycol-water mixture from the seats of h their diaphragm valves (NF-228A, NF-2338, and NF-234A), if 's meeting a zero indicated leakage rate (not including instrument error) for the diaphragm valves. These tests may be used in lieu of tests which are otherwise required by Section III.C.2(a) of 10 CFR 50, Appendix J to use air or nitrogen as

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)(f SURVEILLANCE REOUIREMENTS (Continued)

)

f the test medium.

interval are not changed by this exception.The r.bave required test pre N

e.

Purge supply and exhaust isolation valves with resilient material b

A of Specification 4.6.1.9.3 or 4.6.1.9.4, as applicable; seal 4 f.

k The combined bypass leakage rate shall be determined to be less than h

M 0.07 L, by applicable Type B and C tests at least once per 24 months except for penetrations which are not individually testable;

,. penetrations not individually testable shall be detemined to have s

no detectable leakage when tested with soap bubbles while the con-L tainment each Type A istest; pressurized to P,,14.8 psig. Q'J_p@;t during g.

s

[ Air locks shall4.6.1.3; Specification be tested and demonstrated OPERABLE per 1 h. ,

j The space between each dual-ply bellows assembly on containment t

be vented to the annulus during Type A tests. penetrations betwe

) of each Ty>e A test, the space between each dual-ply bellowsFollowing comple assembly s1all be subjected to a low pressure test at 3-5 psig to i verify no detectable leakage or the dual-ply bellows assembly shall I be subjected to a leak test with the pressure on the containment side of the dual-ply bellows assembly at P ,14.8 psigar7.? P*,

ps4

% to2 f.verify the leakage to be within t$e limits of Specification

4. 6.1.  ;

i

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j All test leakage rates shall be calculated using observed data

converted to absolute values.

select a balanced Integrated Leakage Measurement System; andErro

j. j The provisions of Specification 4.0.2 are not applicable.

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McGUIRE - UNIT 2 3/4 6-4 Amendment No.

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f CONTAINMENT SYSTEMS 2 s

( ) CONTAINMENT VESSEL STRUCTURAL INTEGRITY f

LIMITING CONDITION FOR OPERATI0rf I

3.6.1.6 The structural integrity of the containment vessel shall be maintained 4.6.1.6. at a level consistent with the acceptance criteria in Specification

[ APPLICABILITY: MODES 1, 2, 3, and 4.

S 1 ACTION:

i above requirements, restore the structural integrity to within

% prior to increasing the Reactor Coolant System temperature above 200'F.

fd y -

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SURVEILLANCE REQUIREMENTS k

4.6.1.6 ((%fLR.EAJCs Cp6CIFICRflou 'tG The structural integrity of the containment vessel shall be

( '

4 determined durbn the ehtde.:n for cec'e Type ?. cc;iteiiii. win lealmge.-rmst i 4

(refer-ence-Speci#icatton4.6_1_2} by a visual inspection of the exposed f} accessible interior and exterior surfaces of the vessel. This inspection p shall be performed prior to the Type A containment leadg(" rate" test &to verify 2 3 no apparent changes in appearance of the surfaces or other abnormal l de radation.3 Any abnormal degradation of the containment vessel detected fvi .

j during the~above required inspections shall be reported to the Commission pursuant to 10 CFR Sections 50.72 and 50.73.

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_ REACTOR BUILDING STRUCTURAL INTEGRITY l

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3.6.1.7 p f at a level consistent with the acceptance criteria in Specific

! APPLICABILITY: H0 DES 1, 2, 3, and 4. n ACTION: I lt 1 S  !

h With the structural integrity of the reactor building not conforming to the x! above requirements, restore the structural integrity to within the limits ,

El prior to increasing the Reactor Coolant System temperature above 200*F. g D.

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a McGUIRE - UNIT 2 3/4 6-10 o;:

Amendment No. 148 ?8 4

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' 3/4.6 CONTAINMENT SYSTEMS 9 k

BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY l I Primary CONTAINMENT INTEGRITY ensures that the release of radioactive N materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. This i

h g site boundary radiation doses to within the dose guideline valu Part 100 during accident conditions.

3/4.6.1.2 CONTAINMENT LEAKAGE 4

S The limitations on containment leakage rates ensure that the total i,j R

analyses at the peak accident pressure, P . containment leakage volume w

.. As an added conservatism, the 3 __. measured overall integrated leakage rate ,is .further. limited to less than or 3 equal tests toto 0.75 L,fore"poss account 9.75 'd. n appH+ahl% during performance of the periodic (x between leakage tests. ie degradation of the containment leakage barriers .

}j sW b y j the requirements of Appendix J of 10 CFR 50The surveillance testing for me 3 OPr/0A) O, 'TYF6 6 Mb C 765C 2 b JY.k.$3 N ENT AI K il Mldlb b}' TH l# W $0 A #l0'OblX h 0 0f~l W 0 j

i j required to meet the restrictions on CONTAINMENT INTEGR rate.

y Surveillance testing of the air lock seals provide assurance that the t overall air lock the intervals leakage between willleakage air lock not become tests. excessive due to seal damage- during.

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i McGUIRE - UNIT 2 p{

B 3/4 6-1 I

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N CONTAINMENT SYSTEMS

,i BASES 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that:

1 (1) the

) differential with respect to the outside atmosphere of 1.5 p containment during peak pressure does not exceed the design pressure of 15 psig LOCA conditions.

4 i The maximum peak pressure expected to be obtained from a LOCA event is 14.5 psig.

4 C

wi.11 limit the total pressure to 14.8 psig which is less than th pressure and is consistent with tile accident analyses. i

[o 3/4.6.1.5 AIR TEMPERATURE bI i (1) the containment air mass is limited to an initial massg suffici I g

@ ambient air temperature does not exceed thatthe temperature allo

] continuous within duty rating specified for equipment and instrumentation located containment. d>

d B Measurements shall be made at all listed locations M

P whether temperature. by fixed or portable instruments, prior to determining the avera,ge air il }

C air mass during a LOCA.The containment pressure transient is sensitive to the initially y perature. The contained air mass increases with decreasing tem-y The lower temperature limit of 100*F for the lower compartment, 75*F a for the upper compartment and 60 F when less than or equal to 5's of RATED THERMAL POWER will limit the peak pressure to 14.8 psig which is less than the containment design pressure of 15 psig.

i the peak accident t'emperature 'slightly during a LOCA; however, this limit i based primarily upon equipment protection and anticipated operating conditions $~

i Bothinthe used theupper accidentand lower temperature limits are consistent with the parameters analyses.

F

! I 3/4.6.1.6 i CONTAINMENT VESSEL STRUCTURAL INTEGRITY j

This limitation ensures that the structural integrity of the containment j

t steellife the vessel of thewill be maintained comparable to the original design standardsi for facility. i i vessel will withstand the maximum pressure of 15 psig in the event D<

of a A visual inspection in conjunctier4th Type A leakagc-tat 3 is sufficient to Memonstratethiscapability.

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$a j McGUIRE - UNIT 2 B 3/4 6-2 51 y fy i YS '

it '

CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION l

l 1

3.6.1.2 Containment leakage rates shall be limited to:

a. An overall integrated leakage rate of less than or equal to L ,

0.30% by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at P,,

14.8 psig,

b. A combined leakage rate of less than 0.60 L for all penetrations and valves subject to Type B and C tests, w$en pressurized to P ,

and

c. A combined bypass leakage rate of less than 0.07 L, for all penetra-tions identified as secondary containment bypass leakage paths when pressurized to P,.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With (a) the measured overall integrated containment leakage rate exceeding 0.75 L, or (b) the measured combined leakage rate for all penetrations and valves subject to Types B and C tests exceeding 0.60 L,, or (c) the combined

, restore the overall integrated leakage bypass rate to less than 0.75 L, and the com L,bined leakage rate for all penetrations leakage rate exceeding 0.07 ,

and valves subject to Type B and C tests to less than 0.60 L , and the combined bypass leakage rate to less than 0.07 L, prior to increasing the Reactor Coolant System temperature above 200*F.

SURVEILLANCE REQUIREMENTS 4.6.1.2 Type A containment leakage rates shall be demonstrated as required by 10 CFR 50.54(o) and Appendix J of 10 CFR 50, Option B, as modified by approved exemptions, and in accordance with the guidelines of Regulatory Guide 1.163, September, 1995.

l l

l McGUIRE - UNIT 1 3/4 6-2 Amendment No.

4 CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)  !

l

a. Deleted

b. Deleted

c. The accuracy of each Type A test shall be verified by a supplemental test in accordance with Regulatory Guide 1.163, September,1995.

d. Type B and C tests shall be conducted, in accordance with 10 CFR 50.54(o) and 10 CFR 50 Appendix J, Option A, with gas at P,,

14.8 psig, at intervals no greater than 24 months except for tests involving:

1) Air locks,

2) Dual-ply bellows assemblies on containment penetrations between the containment building and the annulus, and

3) Purge supply and exhaust isolation valves with resilient material seals.

4) Type C tests performed on containment penetrations M372, M373 without draining the glycol-water mixture from the seats of ,

their diaphragm valves (NF-228A, NF-2338, and NF-234A), if I meeting a zero indicated leakage rate (not including instrument error) for the diaphragm valves. These tests may be used in lieu of tests which are otherwise required by Section III.C.2(a) of 10 CFR 50, Appendix J to use air or nitrogen as McGUIRE - UNIT 1 3/4 6-3 Amendment No.

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 3 the test medium. The above required test pressure (P,) and test interval are not changed by this exception.

e.

Purge supply and exhaust isolation valves with resilient material seals shall be tested and demonstrated OPERABLE by the requirements of Specification 4.6.1.9.3 or 4.6.1.9.4, as applicable; f.

The combined bypass leakage rate shall be determined to be less than 0.07 L,for penetrations which are not estable; except individually tby appl penetrations not individually testable shall be determined to have no detectable leakage when tested with soap bubbles while the con-tainment is pressurized to P,,14.8 psig during each Type A test; g.

Air locks shall be tested and demonstrated OPERABLE per Specification 4.6.1.3;

h. The space between each dual-ply bellows assembly on containment penetrations between the containment building and the annulus shall be vented to the annulus during Type A tests. Following completion of each Type A test, the space between each dual-ply bellows j assembly shall be subjected to a low pressure test at 3-5 psig to 1 4

verify no detectable leakage or the dual-ply bellows assembly shall be subjected to a leak test with the pressure on the containment )

side of the dual-ply bellows assembly at P,,14.8 psig to verify the leakage to be within the limits of Specification 4.6.1.2f.; 1 l

i.

All test leakage rates shall be calculated using observed data

-converted to absolute values. Error analyses shall be performed to select a balanced Integrated Leakage Measurement System; and j.

The provisions of Specification 4.0.2 are not applicable.

l

)

McGUIRE - UNIT 1 3/4 6-4 Amendment No.

CdNTAINMENTSYSTEMS CONTAINMENT VESSEL STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.6 The structural integrity of the containment vessel shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With the structural integrity of the containment vessel not conforming to the f above requirements, restore the structural integrity to within the limits i prior to increasing the Reactor Coolant System temperature above 200 F.

SURVEILLANCE REQUIREMENTS

)

4.6.1.6 The structural integrity of the containment vessel shall be I determined by a visual inspection of the exposed accessible interior and  !

exterior surfaces of the vessel. This inspection shall be performed prior to I the Type A containment leakage rate test verify no apparent changes in appearance o(reference Specification 4.6.1.2) to l degradation. f the surfaces or other abnormal If the Type A test is performed at 10-year intervals, two additional inspections shall be performed at approximately equal intervals during shutdowns between Type A tests. Any abnormal degradation of the containment vessel detected during the above required inspections shall be reported to the Commission pursuant to 10 CFR Sections 50.72 and 50.73.

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McGUIRE - UNIT 1 3/4 6-9 Amendment No.

l i

n 1 j C NTAINMENT SYSTEMS REACTOR BUILDING STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 4

3.6.1.7 The structural integrity of the reactor building shall be maintained-at a level consistent with the acceptance criteria in Specification 4.6.1.7.

j APPLICABILITY: MODES 1, 2, 3, and 4. 1 ACTION:

i 5

With the structural integrity of the. reactor building not conforming to the above requirements, restore the structural integrity to within the limits

prior to increasing the Reactor Coolant System temperature above 200*F.

4 SURVEILLANCE REQUIREMENTS l

4.6.1.7 The structural integrity of the reactor building shall be determined during the shutdown for each Type A containment leakage rate test (reference  :

Specification 4.6.1.2) by a visual inspection of the exposed accessible interior and exterior surfaces of the reactor building and verifying no apparent changes in appearance of the concrete surfac2s or other abnormal degradation. If the Type A test is performed at 10-year intervals, two additional inspections shall be performed at approximately equal intervals during shutdowns between Type A tests. Any abnormal degradation of the reactor building detected during the above required inspections shall be reported to the Commission pursuant to 10 CFR Sections 50.72, and 50.73.

l McGUIRE - UNIT 1 3/4 6-10 Amendment No.

3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. This restriction, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the dose guideline values of 10 CFR Part 100 during accident conditions.

3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakaga rates ensure that the total containment leakage volume will not exceed the value assumed in the accident .

analyses at the peak accident pressure, P . As an added conservatism, the I measured overall integrated leakage rate ,is further limited to less than or equal to 0.75 L, during performance of the periodic tests to account for i possible degradation of the containment leakage barriers l between leakage tests. )

The surveillance testing for measuring fype A leakage rates is consistent with the requirements of Appendix J of 10 CFR 50, Option B. Type B and C tests I are conducted in conformance with 10 CFR 50 Appendix J, Option A.

l 3/4.6.1.3 CONTAINMENT AIR LOCKS '

The limitations on closure and leak rate for the containment air locks are required to meet tb restrictions on CONTAINMENT INTEGRITY and containment leak rate. Surveillance testing of the air lock seals provide assurance that the overall air lock lealage will not become excessive due to seal damage during the intervals between air lock leakage tests.

i 4

4 i

McGUIRE - UNIT 1 B 3/4 6-1

CONTAINMENT SYSTEMS BASES 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that: (1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 1.5 psig, and (2) the containment peak LOCA conditions. pressure does not exceed the design pressure of 15 psig during i

The maximum peak pressure expected to be obtained from a LOCA event is 14.5 psig. The limit of 0.3 psig for initial positive containment pressure will limit the total pressure'to 14.8'psig which is less than the design pressure and is consistent with the accident analyses.

3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that:

(1) the containment air mass is limited to an initial mass sufficientl prevent exceeding the design pressure during LOCA conditions, and (2) the y low to ambient air temperature does not exceed that temperature allowable for the continuous duty rating specified for equipment and instrumentation located within containment. Measurements shall be made at all listed locations, whether temperature. by fixed or portable instrumenth prior to determining the average air i l

The containment pressure transient is sensitive to the initially contained air mass during a LOCA. The contained air mass increases with decreasing tem-perature. The lower temperature limit of 100*F for the lower compartment, 75*F for the upper compartment and 60*F when less than or equal to 5% of RATED THERMAL POWER will limit the peak pressure to 14.8 psig which is less than the ,

containment design pressure of 15 psig. The urper temperature limit influences i the peak accident temperature slightly during a LOCA; however, this limit is based primarily upon equipment protection and anticipated operating conditions. I j

Both the upper and lower temperature limits are consistent with the parameters used in the accident analyses.

3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment steel vessel will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that the i vessel will withstand the maximum pressure of 15 psig in the event of a LOCA.

A periodic visual insnection is sufficient to demonstrate this capability.

I McGUIRE - UNIT 1 B 3/4 6-2

CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:

a. An overall integrated leakage rate of less than or equal to L ,

0.30% by weit if the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at P,,

14.8 psig,

b. A combined leakage rate of less than 0.60 L for all penetrations and valves subject to Type B and C tests, w$en pressurized to P ,

and

c. A combined bypass leakage rate of less than 0.07 L, for all penetra-tions identified as secondary containment bypass leakage paths when pressurized to P,.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With (a) the measured overall integrated containment leakage rate exceeding 0.75 L, or (b) the measured combined leakage rate for all penetrations and or (c) the combined valves subject to Types B and C tests exceeding

, restore 0.60 L ,ll integrated leakage the overa bypass leakage rate exceeding rate to less than 0.75 L, and the com 0.07 L,bined leakage rate for all penetrations 1 and valves subject to Type B and C tests to less than 0.60 L,, and the I combined bypass leakage rate to less than 0.07 L, prior to increasing the Reactor Coolant System temperature above 200*F.

SURVEle ;n'E REQUIREMENTS ,

4.6.1.2 Type A containment leakage rates shall be demonstrated as required by 10 CFR 50.54(o) and Appendix J of 10 CFR 50, Option B, as modified by approved exemptions, and in accordance with the guidelines of Regulatory Guide 1.163, Septer.ber, 1995.

hcGUIRE - UNIT 2 3/4 0-? Amendment No.

l t

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 1

a. Deleted

! b. Deleted

c. The accuracy of each Type A test shall be verified by a supplemental test in accordance with Regulatory Guide 1.163, September, 1995.

d. Type B and C tests shall be conducted, in accordance with 10 CFR 50.54(o) and 10 CFR 50 Appendix J,' Option A, with gas at Pa*

14.8 psig, at intervals no greater than 24 months except for tests j involving: -- ~

f 1) Air locks,

2) Dual-ply bellows assemblies on cantainment penetrations between

} the containment building and the annulus, and 1

3) Purge supply and exhaust isolation valves with resilient material seals.

4) Type C tests performed on containment penetrations M372, M373 without draining the glycol-water mixture from the seats of ,

i their diaphragm valves (NF-228A, NF-2338, and NF-234A), if '

meeting a zero indicated leakage rate (not includir.g instrument i error) for the diaphragm valves. These tests may be used in lieu of tests which are otherwise required by Section III.C.2(a) of 10 CFR 50, Appendix J to use air or nitrogen as

! l i

• l i

McGUIRE'- UNIT 2 3/4 6-3 Amendment No.

4 3 CbNTAINMENTSYSTEMS SURVEILLANCE REQUIREMENTS (Continued) _

the test medium. The above required test pressure (P,) and test interval are not changed by this exception.

e. Purge supply and exhaust isolation valves with resilient material I i

seals shall be tested and demonstrated OPERABLE by the requirements I of Specification 4.6.1.9.3 or 4.6.1.9.4, as applicable; f.

The combined bypass leakage rate shall be determined to be less than 0.07 L, by applicable Type B and C tests at least once per 24 months  !

except for penetrations which are not ind:/idually testable; penetrations not individually testable shall be determined to have no detectable leakage when tested with soap bubbles while the con-tainment is pressurized to P,,14.8 psig during each Type A test;

g. Air locks shall be tested and demonstrated OPERABLE per Specification 4.6.1.3;

h. The space between each dual-ply bellows assembly on containment penetrations between the containment building and the annulus shall be vented to the annulus during Type A tests. Following completion of each Type A test, the space between each dual-ply bellows assembly shall be subjected to a low pressure test at 3-5 psig to verify no detectable leakage or the dual-ply bellows assembly shall be subjected to a leak test with the pressure on the containment side of the dual-ply bellows assembly at P ,14.8 psig to verify the  ;

leakage to be within the limits of Specification 4.6.1.2f.;

l i.

All test leakage rates shall be calculated using observed data converted to absolute ilues. Error analyses shall be performed to i select a balanced Integrated Leakage Measurement System; and j

j. i The provisions of Specification 4.0.2 are not applicable.

{

McGUIRE - UNIT 2 3/4 6-4 Amendment No.

1

. 1 CbHTAINMENTSYSTEMS CONTAINMENT VESSEL STRUCTURAL INTEGRITY I

LIMITING CONDITION FOR OPERATION l

3.6.1.6 The structural integrity of the containment vessel shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6.

I APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

Withthestructuralintegrktyofthecontainmentvesselnotconformingto he above requirements, restore the structural integrity to within the limits I prior to increasing the Reactor Coolant System temperature above 200*F.

SURVEILLANCE REQUIREMENTS 4.6.1.6 The structural integrity of the containment vessel shall be determined by a visual inspection of the exposed accessible interior and exterior surfaces of the vessel. This inspection shall be performed prior to the Type A containment leakage rate test (reference Specification 4.6.1.2) to verify no apparent changes in appearance of the surfaces or other abnormal degradation. If the Type A test is performed at 10-year intervals, two additional inspections shall be performed at approximately equal intervals during shutdowns between Type A tests. Any abnormal degradation of the containment vessel detected during the above required inspections shall be reported to the Commission pursuant to 10 CFR Sections 50.72 and 50.73.

McGUIRE - UNIT 2 3/4 6-9 Amendment No.

CONTAINMENT, SYSTEMS l REACTOR BUILDING STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.7 The structural integrity of the reactor building shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.7.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With the structural integrity of the reactor building not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200*F.

1 SURVEILLANCE REQUIREMENTS 4.6.1.7 The structural integrity of the reactor building shall be determined during the shutdown for each . Type A containment leakage rate test (reference 1 Specification 4.6.1.2) by a visual inspection of the exposed accessible interior and exterior surfaces of the reactor building and verifying no apparent changes in appearance of the concrete surfaces or other abnormal degradation. If the Type A test is performed at 10-year intervals, two additional inspections shall be performed at approximately equal intervals

during shutdowns between Type A tests. Any abnormal degradation of the

! reactor building detected during the above required inspections shall be j reported to the Commission pursuant to 10 CFR Sections 50.72, and 50.73.

I 4

1 i

i McGUIRE - UNIT 2 3/4 6-10 Amendment No. l

l .

j 3/4.6 CONTAINMENT SYSTEMS

}

l BASES

3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY i Primary CONTAINMENT INTEGRITY ensures that the release of radioactive

! materials from the containment atmosphere will be restricted to those leakage 4 paths and associated leak rates assumed in the accident analyses. This restriction, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the dose guideline values of 10 CFR Part 100 during accident conditions.

3/4.6.1.2 CONTAINMENT LEAKAGE i

The limitations on containment leakage rates ensure that the total

', containment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure, P . As an added conservatism, the measured overall integrated leakage rate ,is further limited to less than or equal to 0.75 L during performance of the periodic tests to account for ,

possible degrad,ation of the containment leakage barriers l j between leakage tests. l i The surveillance testing for measuring Type A leakage rates is consistent

with the requirements of Appendix J of 10 CFR 50, Option B. Type B and C tests are conducted in conformance with 10 CFR 50 Appendix J, Option A.

3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate. Surveillance testing of the air lock seals 7rovide assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.

i i

i McGUIRE - UNIT 2 B 3/4 6-1

e

• O CONTAINMENT SYSTEMS  !

I BASES 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that: (1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 1.5 psig, and (2) the containment peak pressure does not exceed the design pressure of 15 psig during LOCA conditions.

The maximum peak pressure expected to be obtained from a LOCA event is 14.5 psig. The limit of 0.3 psig for initial positive containment pressure will limit the total pressure-to 14.8 psig which is less than the' design pressure and is consistent with the accident analyses.

3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that:

(1) the containment air mass is limited to an initial mass sufficiently low to prevent exceeding the design pressure during LOCA conditions, and (2) the ambient air temperature does not exceed that temperature allowable for the continuous duty rating specified for equipment and instrumentation located within containment. Measurements shall be made at all listed locations, whether by fixed or portable instruments, prior to determining the average air temperature.

The containment pressure transient is sensitive to the initially contained air mass during a LOCA. The contained air mass increases with decreasing tem-perature. The lower temperature limit of 100*F for the lower compartment, 75'F for the upper compartment and 60*F when less than or equal to 5% of RATED THERMAL POWER will limit the peak pressure to 14.8 psig which is less than the containment design pressure of 15 asig.. The upper . temperature limit influences-- --

the peak accident temperature slig1tly during a LOCA; however, this limit is based primarily upon equipment protection and anticipated operating conditions.

Both the upper and lower temperature limits are consistent with the parameters used in the accident analyses.

3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment steel vessel will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that the '

vessel will withstand the maximum pressure of 15 psig in the event of a LOCA.

A periodic visual inspection is sufficient to demonstrate this capability.

1 McGUIRE - UNIT 2 B 3/4 6-2 i

, ' o o

N 1

e i Attachment II-

, Description of and Justification for Technical Specification Change Description of Changes The changes included in Attachment I will implement the NRC's

revision to 10 CFR 50, Appendix J, which became effective on l October 26, 1995. The revision to the regulation represents a shift away from prescriptive testing requirements in Appendix J, Option A, to a performance-based approach (Option B).

l Specifically, upon completion of two consecutive successful Type A tests, the licensee may extend the test interval up to 10 years

between Type A tests. (Option B also provides for test interval

^

extensions for Type B and C testing, but these changes are not being requested at this time.)

The changes requested herein include:

Specification 3.6.1.2.a.2), which specifies requirements for

reduced-pressure testing, is being deleted. Reduced-pressure testing is not acceptable under the new (Option B) rule.

4 Accordingly, references to reduced-pressure acceptance criteria ,

are also deleted from the ACTION statement. '

Surveillance Requirement 4.6.1.2 is being revised to refer to the requirement in 10 CFR 50.54 (o) that containment testing be ,

performed pursuant to Appendix J; the reference to Appendix J l that currently exists in SR 4.6.1.2 will now refer to Option B of the Appendix; a reference to Regulatory Guide 1.163, September, 1995, is being added. RG 1.163 is the implementation document for

! the new rule.

! SR 4.6.1.2.a and b. are deleted. The test schedule is now 4 determined based upon the criteria of the implementing documents.

In SR 4.6.1.2.c, a reference is added to RG 1.163, dated September, 1995, as the implementing document; and redundant and/or obsolete requirements (c.1), 2), and 3)) are deleted.

Surveillance requirements 4.6.1.6 and 4.6.1.7, which specify periodic inspection requirements for the containment vessel and reactor building, respectively, are also being changed. The inspection frequencies are linked to the performance of the CLRT; increasing the interval of this test would have the unintended (and unacceptable) effect of extending the interval of the 1

containment vessel and reactor building inspections. The I specified inspection intervals are consistent with Appendix J, Option B.

A change to BASIS 3/4.6.1.2 specifies that the as-left containment leakage shall be less than or equal to . 7 5 L. , to account for possible degradation of the containment between tests. Also, a reference to a reduced-pressure test criterion was deleted, and a reference to Option B of Appendix J was added.

Technical Justification The proposed changes are based on approved guidance documents from the NRC and Nuclear Energy Institute (NEI), including NEI 94-01, dated July 26, 1995; Regulatory Guide 1.163, dated September, 1995; and sample Improved Standard Technical Specifications (ISTS) developed by NEI, with NRC cooperation.

The sample ISTS provided guidance on the scope of changes that the NRC expects to see from each of the utilities who elect to pursue Option B. The changes presented in this application meet the intent of the changes, relative to Type A testing, that have been approved in concept by the NRC. The NRC has determined that the industry guideline (NEI 94-01) referenced in the Regulatory Guide, with some exceptions, is an acceptable means of demonstrating compliance with the requirements of Option B. Duke Power intends to comply with the provisions of the NEI document, except as modified by the Regulatory Guide.

The as-found acceptance criterion for Type A tests, L., as specified in TS 3.6.1.2, has not changed, nor has the requirement that the containment leakage be less than or equal to .75 L.

before entering a mode in which containment integrity is required.

Deleting the details of the test program from TSs, and providing a reference to the guidance document (RG 1.163) is consistent with the recommendations of the Regulatory Guide.

The change in the test interval, based on the performance of the l containment structure in previous tests, has been determined by I the NRC's own analysis, presented in NUREG-1493, to have a minimal impact on safety.

The proposed changes will not require revisions to McGuire's UFSAR.

2

~ , ,

e l

I Attachment III No Significant Hazards Analysis The following analysis is presented, pursuant to 10 CFR 50.91, to demonstrate that the proposed change will not create a Significant Hazard Consideration.

1. The proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated.

Containment leak rate testing is not an initiator of any accident; the proposed change does not affect reactor operations or accident analysis, and has no significant radiological consequences. Reliability of the structure Therefore, this proposed change will not involve an increase in the probability or consequences of any previously-evaluated accident.

2. The proposed change will not create the possibility of any new accident not previously evaluated.

The proposed change does not affect normal plant operations or configuration, nor does it affect leak rate test methods. The test history at McGuire (two consecutive successful tests) provides continued assurance of the leak tightness of the containment structure.

3. There is no significant reduction in a margin of safety.

The proposed changes are based on NRC-accepted provisions, and maintain necessary levels of reliability of containment integrity. The performance-based approach to leakage rate testing ,

recognizes that historically good results of containment testing l provide appropriate assurance of future containment integrity; l this supports the conclusion that the impact on the health and safety of the public as a result of extended test intervals is negligible. In addition, local leakrate testing will continue to provide assurances of overall containment integrity.

Based on the above, no significant hazards consideration is created by the proposed change.

3

~ , . -

,. .e Attachment III, continued Environmental Assessment The proposed change has been evaluated to determine if any environmental impact would be created. The change is considered to meet the criteria (presented in 10 CFR 51.22 (c) (9) ) for categorical exclusion from the requirements for an environmental assessment, because:

A. As documented above, the change will create No Significant Hazards Consideration.

B. There is no change in the type, or significant increase in the amounts, of any effluent that may be released offsite.

The change will create no new mechanism by which effluents are released, and will provide continued assurances that leakage remains within the existing allowed leakage, L.,

C. There is no significant increase in individual or cumulative occupational radiation exposure.

The proposed change will not change methods by which radioactive materials, including effluents, are handled, processed, or disposed of. Normal radiation levels within the nuclear station will not increase, and this change will not result in personnel spending additional time in radiation areas.

Therefore, there will be no increase in individual or cumulative radiation exposure.

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