ML20096D446

Proposed Tech Specs,Adopting Option B of 10CFR50,App J to Require Type a Containment Lrt to Be Performed on performance-based Testing Schedule

ML20096D446
Person / Time
Site:	Calvert Cliffs
Issue date:	01/16/1996
From:	BALTIMORE GAS & ELECTRIC CO.
To:
Shared Package
ML20096D432	List: ML20096D429 ML20096D446
References
NUDOCS 9601190134
Download: ML20096D446 (21)

Text

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3/4.6 CONTAINMENT SYSTEMS

, 3/4.6.1 PRIMARY CONTAINMENT Containment Leakaae i

i LIMITING CONDITION FOR OPERATION

! 3.6.1.2 Containment leakage rates shall be. limited to: TEr4 ace Wir#

2 i a. M ever:ll integr:ted 1::k g: r:t: cfr INSE4 i

4. 2 L. (ne,000 scen), 0.20 3:r= t by =ight f th:

= t:in nt :f r p:r 21 heer: :t P., 50 p:is, er

/

4

4. 2 'g -(*51,000 S00"), 0.050 per::;t by =ight of the per 21 5:cr: :t : 7:d =d prn:;r: ef P ,

-25 peig. - e A -

b. A combined leakage rate of 5 0.50 L (173,000 SCCM), for all penetrations and valves subject to type B and C tests when l

pressurized to P..

de etccephnnce c.eiferia. spesiRed in de.

APPLICABILITY: MODES 1, 2, 3 and 4. Conb41 ament

~ leakaye. Wafe 7Es//ng 3*) fem > i

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ACTION: With either (a as ed ove 11 inte te i l g o ay ==C a0_v.75 % w " . ;_C" .. . ,5 y ('.,2.,SC_..),:.j i s p...__.:V or tb) with the measured combine eakage rate for all.

penetrationsandvalvessubjecttoTypesBandCtestsexceedino0.504, restore th containme lesi _then er; Qi t; . 75 ' er . ; ...= Or ;;5: L .. ,. 5 ' , E :;;;1 [:b, 1 Ji and the combined leakage rate or all penetrations and valves subject to Type B and C tests to less than or equal to 0.50 L, prior to increasing the Reactor Coolant System temperature above 200 F.

WiUin Ehe Acceptance criteria. specih'ed in de Containment .

Leakage %fe Te~stin Progrann SURVEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria, methods and provisions specified in 10 CFR Part 50, Appendix J:

e

a. Three Type A tests (Overell !atograted Cen ainant Leakage note) shell be n ,d;;ted :t :pp =i=tely Oge:1 interv:1: during-

-hatd;.;r. :t either P

- ,o q ..y yo, ___

..______m__

L(50iy_ p:ig) ervat P, (25 p:ig) ~

during :::h-Perf.ru tapa;ted tesfiny in accoedance with de Centainment leakage " &fe 7EsthO lbgnam ,

CALVERT CLIFFS - UNIT 1 3/4 6-2 Amendment No. 169 9601190134 960116 PDR ADOCK 05000317 p PDR

_ . _ _ . . . _ . . _ - . _ _ . _ . _ _ _ _ . _ . _ _ _ . _ _ _ . _ . _ . . . . . . _ _ _ . _ . .. _ _._.m . - _ . _ _ _ _ . _ . . - . . . _

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INSERT A  !

a. A maximum allowable containment leakage rate, La, as specified in the Specification 6.19,-

" Containment Leakage Rate Testing Program."

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8 3/4.6 CONTAINMENT SYST M OEL ETE SURVEILLANCE REQUIREMENTS (Continued) i _

. I any rio c Ty At f 1s me t ei er .75 (2 500 CCM) or 0. 5 L, 6 S ), he t st hed.le r sub quen Type tes s s 1 rev nd pro db th Com sion. If wo co secui, ve pe tes fa to et eit r

.75 (259 00 CM) 0. L, 6,2 SC ). Typ A t st s 11 per med at le st a no hs til wo nse ti Ty A t sts et e ther .75 ry(2 500 SC or .75 (46,00S )a whi ti the boy tes sch ule y s .

. T acc racy f ea Ty A t st s 11 ve fie by su leme tal t st w ich:

. C nfi the accu cy the Type te by eri ing hat th dif renc bet en nta and te da is wit n 0. 5 L, 86, S pp)lo 0.25 L. ( ype4 SC .

. sa urat ns fic nt est blis acc ate the cha e

.i leak eb wee the ype tes and upp n 1 te t.

3. Requ res eq ntit of as ect d in th cont inme t o led om e co tai nt urin the supp n 1 te t to be e iva nt t at east of he t al asu d le age ate t P, 50 5igp) rcenr P. (25 sig).

Type B andI tests sliall be conducted with gas at P (50 psig) at s intervals of 24 months except for tests involving air locks.

Ju(. Air locks shall be tested and demonstrated OPERABLE per Surve 1 Requirement 4.6.1.3.

' X All tes ealuFge rates shall be calculated using observed cata converted _ to absolute =:1y::: :n11 h grf ...d i n ig . . . . ..,. . . . _, . _. .

. Containment purge isolation valves shall be demonstrated OPERABLE

@'w any time upon entering M0DE 5 from power operation modes, unless the last surveillance test has been perfomed within the past 6 months or any time after being opened and prior to entering M0DE 4 from shutdown modes by verifying that when the measured Exemption to 10 CFR Part 50, Appendix J.

CALVERT CLIFFS - UNIT 1 3/4 6-3 Amendment No. 169

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, 3/4.6 CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) g i leakage rate is added to the le <

rates determined pursuant to Technical Specification 4.6.1.2 sfor all other Type B or C j penetrations, the combined leakage rate is less than or equal to 0.50 L 173,000 SCCM). The leakage rate for the containment 2

purge ls(olation valves shall also be compared to the previousl measured leakage rate to detect excessive valve degradation.

' . The containment purge isolation valve seals shall be replaced  ;

with new seals at a frequency to ensure no individual seal remains in service greater than 2 consecutive fuel reload cycles. i f

CALVERT CLIFFS - UNIT 1 3/4 6-4 Amendment No. 169

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3/4.6 CONTAINMENT SYSTEMS 1

SURVEILLANCE REQUIREMENTS (Continued) 4.6.1.6.2 End Anchoraces and Ad.iacent Concrete Surfaces. The structural integrity of the end anchorages and adjacent concrete surfaces shall be demonstrated by detemining through inspection of a representative sample of tendons (reference Specification 4.6.1.6.1) that no apparent changes have occurred in the visual appearance of the end anchorages or their I adjacent concrete exterior surfaces. Also, inspections of the pre-selected i concrete crack patterns adjacent to end anchorages shall be perfomed during the Type A containment leakage rate tests (reference Specification 4.6.1.2) while the containment is at its maximum test pressure.

4.6.1.6.3 Containment Surfaces. The exposed accessible interior and exterior surfaces of the containment includina the liner plate shall be visually i e d ur ng He sh o f ch Ty~p A on i n7 ka a ~ts (re er nc Sp ci ica io 4. 1. . Th i spe ti s al be er o d rio t te pe c nt in. t a ge at te t u o r y id nc o st et r d ter or to wh ch y ff t ith r e ,

4 c t n nt t et a in egr ty r eak tig tn s. l 4.6.1.6.4 . Reports. Any abnormal degradation of the containment structure detected during the above required tests and inspections shall be reported to the Consission pursuant to Specification 6.9.2 within the next 30 days.

This report shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, and the corrective actions taken.

_ ~ v v v in accordance S with de Contain,,,,,J Leakage ~7Rate visL %Q (teference ycl0cabjon Y.

CALVERT CLIFFS - UNIT 1 3/4 6-11 Amendment No. 169

6.0 ADMINISTRATIVE CONTR0LS

( 6.17.2 Licensee initiated ~ changes to the ODCM:

a. Shall be submitted to the Commission in the Semiannual .

Radioactive Effluent Release Report for the period in which the l change (s) was made effective. This submittal shall contain:

1. . Sufficient information to support the rationale for the  !

change. Infomation submitted should consist of a package of those pages of the ODCM to be changed with each page numbered  ;

and provided with a change number and/or change date together with appropriate analyses or evaluations justifying the change (s);

2. A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint deteminations; and

3. Documentation of the fact that the change has been reviewed and found acceptable by the POSRC.

b. Shall become effective upon review by the POSRC and approval of the Plant General. Manager.-

6.18 MAJOR CHANGES TO RADI0 ACTIVE LIOUID. GASEOUS AND SOLID WASTE TREATM ENT SYSTEMS 6.18.1 Licensee initiated major changes to the Radioactive Waste Systems (liquid, gaseous and solid) shall be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the modification to the waste system is completed. The discussion of each change shall contain:

a. A description of the equipment, components and processes involved.

b. Documentation of the fact that the change including the safety analysis was reviewed and found acceptable by the POSRC.

INSEAT* b i

CALVERT CLIFFS - UNIT 1 6-31 Amendment No. 186 l

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(NOTE: This Specification will be renumbered 6.5.6 following approval of BGE's Administrative Controls License Amendment Request, dated March 15,1995. Other references to Specification 6.19 will also be changed to 6.5.6 following approval of that License Amendment Request.)

INSERT B 6.19 Contamment Leakage Rate Testing Program A program shall be established to implement the leakage testing of the containment as required by 10 CFR 50.54 (o) and 10 CFR Part 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, " Performance-Based Contamment Leak-Test Program," dated September 1995, as modified by approved exceptions.

The peak calculated containment internal pressure for the design basis loss-of-coolant accident, Pa.

is 49.4 psig. The containment design pressure is 50 psig.

The maximum allowable containment leakage rate, L a , shall be 0.20 percent of containment air weight per day at Pa-Containment leakage rate acceptance criterion is s 1.0aL . During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria is s 0.75 aL for Type A tests.

The provisions of Specification 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of Specification 4.0.3 are applicable to the Containment Leakage Rate Testing Program.

.. i 3/4.6 CONTAIMENT SYSTEMS  !

i BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY In MODES 1, 2, 3, and 4, primary CONTAIMENT 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 i accident analyses. This restriction, in conjunction with the leakage rate i limitation, will limit the SITE B0UNDARY radiation doses to within the limits of 10 CFR Part 100 during accident conditions. In MODES 5 and 6,  ;

the probability and consequences of these events are reduced because of the Reactor Coolant System (RCS) pressure and temperature limitations of these modes, by preventing operations which could lead to a need for containment isolation, and by providing containment isolation through penetration closure. .

1 k/ITH WSMr 0 3/ 6.1. Chai [bEPLACE co at 11 tat ns\ocont inmewi lenotagexceetes t1 kage vol nsur tha th tot 1 the alue ss d t cci nt alys at he p ka iden pre ure, %. a ad d c se tis the as ed o eral int rate lea: ge r e fu the 11 ted 5 75 or _ ( ap icab )d ing rfo anc of he eri ic t ts ac unt or p sib de adat no the nt nme t ,

1 kag bar ers etwe le age ests.

es vei anc tes ng r suri le age ates re nsis ent wit th re ir ts 10 FR rt . Ap endi J, cep for ep fo anc f e and lea ge sti . ea owa e le kage ate as b en i p po ion ely edu d re n ed i Gen ic tte 91-0 to cc nt for n ten ds vei R 0.2 anca sch ule f 24 nt 9 + (r  !

I ec ica on Ap nd J. *(s i Th an cep on omKCF Part 50 l 3/4.6.1.3 Containment Air Locks l

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 provides assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.

i CALVERT CLIFFS - UNIT 1 B 3/4 6-1 Amendment No. 169

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.. .i INSERT C 3/4.6.1.2 Containment Leakane Maintaining the containment OPERABLE requires compliance with the visual examinations and leakage rate test requirements of the Containment Leakage Rate Testing Program for Type A tests, and 10 CFR Part 50, Appendix J, Option A for Type B and C tests. As-left leakage prior to the first startup after performing a required leakage test is required to be s 0.75 L. for overall Type A leakage. At all other times between required leakage rate tests, the acceptance criteria is based on an overall Type A leakage limit of s 1.0 L.. At s 1.0 L the offsite dose consequences are bounded by the assumptions of the safety analysis. 'Ihe frequency of Type A testmg is specified in the Containment Leakage Rate Testing Program.

'Ihe surveillance testing for measuring leakage rates are consistent with the requirements of 10 CFR Part 50, Appendix J, Option B for Type A tests. For Type B and C testing, the allowable leakage rate has been proportionately reduced, as recommended in Generic Letter 91-04, to account for an extended surveillance schedule of 24 months + 25% (per Specification 4.0.2). This is an exception from 10 CFR Part 50, Appendix J.

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3/4.6 CONTAINMENT SYSTDt$

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3/4.6.1.4 Internal Pressure i

The limitations on containment internal pressure ensure that 1) the

! containment structure is prevented from exceeding it design negative i pressure differential with respect to the outside atmosphere of 3.0 psig and 2) the containment peak pressure does not exceed the design pressure of l 50 psig during LOCA or steam line break conditions.

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] The maximum peak pressure expected to be obtained from a LOCA event is 47.6 l psig assuming an initial containment pressure of 14.7 psia. The limit of

1.8 psig for initial positive containment pressure will limit the total pressure to 49.4 psig which is less than the design pressure and is

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4 consistent with the accident analyses. The maximum peak pressure expected to be obtained from a steam line break event is 49.2 psig assuming an

initial containment pressure of 16.5 psia (1.8 psig).

3/4.6.1.5 Air Temocrature j The limitation on containment average air temperature ensures that the containment peak air temperature does not exceed the design temperature of 4

1 276 F during LOCA conditions. The containment temperature limit is consistent with the accident analyses.

f 3/4.6.1.6 Containment Structural Intearity i

The limitation ensures that the structural integrity of the containment

{ will be maintained comparable to the original design standards for the life i i

j of the facility. Structural integrity is required to ensure that the containment will withstand the maximum pressure of 47.6 psig in the event 1

i of a LOCA. The measurement of containment tendon lift off force, the visual and metallurgical examination of tendons, anchorages and liner and i the Type A leakage tests are sufficient to demonstrate this capability.

I The surveillance requirements for demonstrating the containment's

! structural integrity are consistent with the intent of the reconmendations

! of Regulatory Guide 1.-35 " Inservice Surveillance of Ungrouted Tendons in j Prestressed Concrete Containment Structures". January 1976.

The end anchorage concrete exterior surfaces are checked visually for indications of abnormal material behavior during tendon surveillance.

Inspections of pre-selected concrete crack patterns are perfomed during the Type A containment leakage rate tests, consistent with the Structural Integrity Test.,

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] f Yisaa,I Inspections of lbe accessible infeti*t ud ederier sur)$aes of fbe con fajn,,,ent are perhemed to allm &t ese) uncovering of evidence Ava  :

n .9 st~.tu,ai .teka.nas.n. # rr.p.ncy .r >Aasa ins, e.ne is in 1

ape,,,ad wilf Ap ahm l Gde 1.145, dded syk,,k MS A - A CALVERT CLIFFS - UNIT 1 B 3/4 6-2 Amendment No. 169

ATTACHMENT (4)

UNIT 2 MARKED-UP TECHNICAL SPECIFICATION PAGES 3/4 6-2 3/4 6-3 3/46-4 3/4 6-10 6-31 B 3/4 6-1 B 3/4 6-2 l

Ealtimore Gas a id Electric Company License Ame;:dment Request January 16,1996

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Leakaae LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to: REmcE wrW Ov:=11 int ;nt d 10:h:;: nt: Of: INSEAT A

-1.  ?? 000 000"), 0.20 p: n;at by ::isht Of th:

2:: 'E.:(ic ;;t mir ;:r 24 b-5 t P., 50 p:!;. er

-2.

... .-- _... . . . . u, S.CL".,).,.

2.._'".1.,(,44,5^^

6.. . ..

.+.w....a._r_.......,

0.^'2 ;;r:: t by ::!;ht of th:

. . -- y

b. A combined leakage rate of < 0.50 L (173,000 SCCM), for all penetrations and valves subTect to type B and C tests when pressurized to P .

be accepfance criferia. specibeol in lbs APPLICABILITY: MODES 1, 2, 3 and 4.

CeJay.n dleakage hte 7isk,,g % yam, ACTION: With either j grate aufadind(a[g,overall ,._,,s 2 .,,, ingcontainment$_001. Or 0.75, J33,400

,. . ... .: or to) w1T.n the niiasured combined leakage rate for all' pene rat ons and valves subject to Types B and C tests exceo. ding i 0.50 L ,

Tht: w C;5;l30 f.7"7 -f 10:t tatetr.!r.for f:q;;;rall 10 penetrations 0~.Z5 4,_:: and n valves e

subl c6mbinedleaia o Type B and C tests to less than or equal to 0.50 L, prior to increasing the Reactor i' Coolant System temperature above 200 F.

j WY In Ebe accepbtnce cdkvia specIAodin ] e Con %ed leahap 2?afe Tsh,9 % gen,n ,

SURVEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the i following test schedule and shall be determined in confomance with the criteria, methods and provisions specified in 10 CFR Part 50 Appendix J:

a. Ty ^ t::t:

b ;;:: d::t:d derta; :h tdr. :t either P. (50 p;ig, e-et-(0=n11 Ir.t:;nte) l P (25 p;ig). The in gt:r.:y Of the Ty p A t::t: :h:11 be ir,J

/:: e=dence with 10 CFR Pert 50. Appen M L es =a m i ,4 by app- :d e--ti:::.

1 Nerin uired visaal exantinkiens anol 5pe A fedin in accordance with lbe entainme CALVERT CLIFFS - UNIT 2 3/4 6-2 Amendment No. 183

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INSERT A i

a. A maximum allowable containment leakage rate, La, as specified in the Specification 6.19, i

" Containment Leakage Rate Testing Program."

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3/4.6 CONTAllMENT SYSTEMS it SURVEILUDICE REQUIREMENTS (Continued) 3 i I.'

b any eri ic T eA est ils o me eit r 0. L 4

( 9,5 SCCM or 75 L (33, 00 S ), e te t se dtIlefor

. sub quen Typ A te ts s 11 rey ed d ap rove by t e

Comm sion. If c sec ive pe A test fail o me t ei er

.75 L. (259 00 CM) 0. L, 3,4 SC ,a pe A test s 11 b per d t le te son s un 1t con cut er ry(25 r 0.

Ty (33, At ts t ei 75 L 500 CM) L, ek j SC ) at whic time the ove est hedu e ma be i es .

3 Th accu cy o each Type tes sha be eri db a i supp nt 1 tes whi :

1 Con ims he a urac of t e Ty e A st ver fyin that the ffer' ce twee supp n 1a Typ At t da is ithin 0.25 (8 500 CM) 0. L, 11,1 SC ).

Ha a du tion uffi ent o es bli ac rate y th chan in 1 kage etwe th Type tes an supp emen. 1 te t.

. R uire the uant of as i ect in th cont inme or bl f the ntai nt urin the -Jpp nt 1 te to e equi lent to at leas to 1 sure rat at P. (50 25igp) cent of5pt g). -

aka P.

(

Type B and C tests shall be conducted with gas at P (50 psig) at intervals of 24 months except for tests involving air locks.

C. . Air locks shall be tested and demonstrated OPERABLE per Surve ance equirement 4.6.1.3.

cd. .

All convertedte ^Teakage to absolute'value rates shall be calculated usingteobserveh.

r.:ly :: f;;11 n . era j

1 f.:Q:yed 1::k;;;; 2 ;;r;nr.t :ystag;;;.

. Containment purge isolation valves shall be demonstrated OPERABLE any time upon entering M0DE 5 from POWER OPERATION MODES, unless the last surveillance test has been perfomed within the past 6 months or any time after being opened and prior to entering MODE 4 from shutdown modes by verifying that when the measured t

Exemption to 10 CFR Part 50, Appendix J.

CALVERT CLIFFS - UNIT 2 3/4 6-3 Amendment No. 149

4

.*. * , s l 3/4.6 CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) h )

leakage rate is added to the lea a rates determined pursuant to Technical Specification 4.6.1.2 or all other Type B or C penetrations, the combined leakage rate is less than or equal to 0.50 L (173,000 SCCM). The leakage rate for the containment purge isolation valves shall also be compared to the previously measured leakage rate to detect excessive valve degradation.  !

{, The containment purge isolation valve seals shall be replaced with new seals at a frequency to ensure no individual seal remains in service greater than 2 consecutive fuel reload cycles.

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CALVERT CLIFFS - UNIT 2 3/4 6-4 Amendment No. 149

3/4.6 CONTAINMENT SYSTEMS 4

SURVEILLANCE REQUIREMENTS (Continued) l 4.6.1.6.3 Containment Surfaces. The exposed accessible' interior lan l exterior surfaces of the containment including the liner late shall bel I VLsually insne ed u in t u o r ch yp A o ai e ae fite e e r nc S ec fi at on 4. 1. ). Th i sp t n ha 1 e se fo p o t te yp A co ta nt lea ge at t t o ne ve ay Ivi en e f t et ra t to at o wh ch ay ff t it r he ,

l c nt_ i n s ru t al i e it ak ti htn ss. '

4.6.1.6.4 Reports. Any abnomal degradation of the containment structure detected during the above required tests and inspections shall be reported to the Consnission pursuant to Specification 6.9.2 within the next 30 days. ,

This report shall include a description of the _ tendon condition, the '

condition of the concrete (especially at tendon anchorages), the inspection  ;

procedures, the tolerances on cracking, and the corrective actions taken. 1 in accotNance / ConlainXd les h am gf (referey Speeihedian Y.4.V l

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CALVERT CLIFFS - UNIT 2 3/4 6-10 Amendment No. 149

6.0 ADMINISTRATIVE CONTROLS b 6.17 0FFSITE DOSE CALCULATION MANUAL (0DCM) 6.17.1 The ODCM shall be approved by the Comission prior to implementation.

6.17.2 Licensee initiated changes to the ODCM:

a. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made effective. This submittal shall contain:

1. Sufficient information to support the rationale for the change. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with a change number and/or change date together with appropriate analyses or evaluations justifying the change (s);

2. A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and

3. Documentation of the fact that the change has been reviewed

, and found acceptable by the POSRC.

b. Shall become effective upon review by the POSRC and approval of

'( the Plant General Manager.

6.18 MAJOR CHANGES TO RADI0 ACTIVE LIQUID. GASE0US AND SOLID WASTE TREATMENT SYSTEMS 6.18.1 Licensee initiated major changes to the Radioactive Waste Systems (liquid, gaseous and solid) shall be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the modification to the waste system is completed. The discussion of each change shall contain:

a. A description of the equipment, components and processes involved,

b. Documentation of the fact that the change including the safety

, analysis was reviewed and found acceptable by the POSRC.

Inssxr 8

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CALVERT CLIFFS - UNIT 2 6-31 Amendment No. 163 l

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.s ,- ,

l (NOTE: This Specification will be renumbered 6.5.6 following approval of BGE's Administrative l Controis License Amendment Request, dated March 15,1995. Other references to Specification 6.19 will also be changed to 6.5.6 following approval of that License Amendment Request.)

INSERT B 6.19 Containment Leakage Rate Testing Program A program shall be established to implement the leakage testing of the containment as required by 10 CFR 50.54 (o) and 10 CFR Part 50, Appendix J, Option B. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, " Performance-Based l Containment Leak-Test Program," dated September 1995.

The peak calculated containment internal pressure for the design basis loss-of-coolant accident, Pa.

is 49.4 psig. The containment design pressure is 50 psig.

The maximum allowable containment leakage rate, La , shall be 0.20 percent of containment air weight per day at Pa-  ;

Containment leakage rate acceptance criterion is s 1.0aL . During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria is s 0.75 aL for Type A tests.

He provisions of Specification 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

He provisions of Specification 4.0.3 are applicable to the Containment Leakage Rate Testing Program.

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3/4.6 CONTAINMENT SYSTEMS i RASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY In MODES 1, 2, 3, and 4, primary CONTAIMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be l restricted to those leakage paths and associated leak rates assumed in the l accident analyses. This restriction, in conjunction with the leakage rate I limitation, will limit the SITE B0UNDARY radiation doses to within the limits of 10 CFR Part 100 during accident conditions. In MODES 5 and 6, the probability and consequences of these events are reduced because of the '

Reactor Coolant System (RCS) pressure and temperature limitations of these modes, by preventing operations which could lead to a need for containment isolation, and by providing containment isolation through penetration RgnAcs wirn fasEA'r 0 3 .6. 2 nta nt Leak e

( he mit ion on c ta nt eak e r tes nsu th th to 1 nt leak ey ume ill ot e cee the valu ass (cactai dent ana ses t th pea acc ent res ure P.. As d nte ad d ons vat m, e sur ov all nteg ate lea age ate s th 1 ite to _ 0. L, 5 75 (as ppl ab )d in er rma e

{pe odi tes to cco tf po ible egr dat no th con i nt eak eb rie be ween leak et ts.

The ur ill ce sti fo mea rin lea ge te are con ste t w h e gaiettan o O RP t5 App dix , ep fo th per rma ce j to Typ BadC eaka e t tin T al wab le ag ra ha be

' ro rt at r uce as ec n in ene c L tte 91 4, o a cou ra ex nde sur ill ce che le 24 nth +  % er p cif ati 4. 2). Th is n e empt nt 10 RP t 0, pe ix .

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 provides assurance I that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.

i CALVERT CLIFFS - UNIT 2 B 3/4 6-1 Amendment No. 149

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INSERT C 3/4.6.1.2 Containment Leakane Maintaining the containment OPERABLE requires compliance with the visual examinations and leakage rate test requirements of the Containment Leakage Rate Testing Program for Type A tests, and 10 CFR Part 50, Appendix J, Option A for Type B and C tests. As-left leakage prior to the first startup after performing a required leakage test is required to be s 0.75 L. for overall Type A leakage. At all other times between required leakage rate tests, the acceptance criteria is based on an overall Type A leakage limit of s 1.0 L.. At s 1.0 L. the offsite dose consequences are bounded by the assumptions of the safety analysis. 'Ihe frequency of Type A testing is specified in the Containment Leakage Rate Testing Program.

The surveillance testing for measuring leakage rates are consistent with the requirements of 10 CFR Part 50, Appendix J, Option B for Type A tests. For Type B and C testing, the allowable leakage rate has been proportionately reduced, as recommended in Generic Letter 91-04, to account for an extended surveillance schedule of 24 months + 25% (per Specification 4.0.2). This is an exception from 10 CFR Part 50, Appendix J.

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3/4.6 CONTAIlptENT SYSTEMS f

usEs 3/4.6.1.4 Internal Pressure The limitations on containment internal pressure ensure that 1) the containment structure is prevented from exceeding it design negative pressure differential with respect to the outside atmosphere of 3.0 psig and 2) the containment peak pressure does not exceed the design pressure of 50 psig during LOCA or steam line break conditions.

The maximum peak pressure expected to be obtained from a LOCA event is 47.6 psig assuming an initial containment pressure of 14.7 psia. The limit of 1.8 psig for initial positive containment pressure will limit the total pressure to 49.4 psig which is less than the design pressure and is consistent with the accident analyses. The maximum peak pressure expected to be obtained from a steam line break event is 49.2 psig assuming an initial containment pressure of 16.5 psia (1.8 psig).

3/4.6.1.5 Air Temocrature The limitation on containment average air temperature ensures that the containment peak air temperature does not exceed the design temperature of 276'F during LOCA conditions. The containment temperature limit is censistent with the accident analyses.

l 3/4.6.1.6 Containment Structural Intecrity '

The limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that the containment will withstand the maximum pressure of 47.6 psig in the event of a LOCA. The measurement of containment tendon lift off force,-the visual and metallurgical examination of tendons, anchorages and liner and the Type A leakage tests are sufficient to demonstrate this capability.

The surveillance requirements for demonstrating the containment's structural integrity are consistent with the intent of the recommendations  !

of Regulatory Guide 1.35 " Inservice Surveillance of Ungrouted Tendons in '

Prestressed Concrete Containment Structures", January 1976.

The end anchorage concrete exterior surfaces are checked visually for indications of abnormal material behavior during tendon surveillance. 3 Inspections of pre-selected concrete crack patterns are performed during the Type A containment leakage rate tests, consistent with the Structural Integrity Test. .

Visual inspeelians of be acca.ssible mferior and exkrier . surfaces of conb;nmeni m perf,,4 h allow F., emely uneeva,;ng d evidanee ,9 hov- = stulwal Jakieuhan. 7Ce fneguency .P dese inspachons is in 1

ayeement wdA %ulahm Saata C163, lated September 1995.

A M-CALVERT CLIFFS - UNIT 2 B 3/4 6-2 Amendment No. 149