Text

Proposed Tech Specs,Removing Prescriptive Requirements of 10CFR50,Option a & Replacing Requirements W/Requirements in Option B
	ML20095G057
Person / Time
Site:	Beaver Valley
Issue date:	12/15/1995
From:	; DUQUESNE LIGHT CO.
To:	;
Shared Package
ML20095G056	List: ML20095G053; ML20095G057;
References
	NUDOCS 9512200004
	Download: ML20095G057 (70)
	v • d • e

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ATTACHMENT A-1 Beaver Valley Power Station, Unit No. 1 i Proposed Technical Specification Change No. 223 i I

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The following is a list of the affected pages:

) Affected Pages: XVII I i 3/4 6-1

, 3/4 6-2 3/4 6-3 1

3/4 6-4

, 3/4 6-5 3/4 6-5b

! 3/4 6-Sc l 3/4 6-10 i

J B 3/4 6-1

B 3/4 6-2 i B 3/4 6-4 l

! B 3/4 6-5 1

{ B 3/4 6-6 l

} B 3/4 6-7 )

] B 3/4 6-8 1 B 3/4 6-9

6-20 6-25 1

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9512200004 951215 PDR ADOCK 05000334 j p PDR i

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DPR-66

. INDEX 1

ADMINISTRATIVE CONTROLS j 4

SECTION PAGE 6.9.1.10' Annual Radiological Environmental Operating j Report. . ................ 18 4

6.9.1.11 Annual Radioactive Effluent Release Report. . ................ 6-19 6.9.1.12 Core Operating Limits Report. ...... 6-19 6.9.2 SPECIAL REPORTS . . . . . . . ...... 6-20 4 .

6.10 RECORD RETENTION. ................ 6-21 1 6.11 RADIATION PROTECTION PROGRAM. . . . . ...... 6-22 i 6.12 HIGH RADIATION AREA ............... 6-23 1

3 j 6.13 PROCESS CONTROL PROGRAM (PCP) . . . . ...... 6-24 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM). ..... 6-24 6.16 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS ................ 6-:5 6a1 CoNTA1MMN1 LEtM ACn # ATE TES7ING- PR06 R AM ..., C-95 l A00 i l

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BEAVER VALLEY - UNIT 1 XVII Amendment No. tee--

(fropotec}Webu) g

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DFR-66 3/4.6 CONTAINMENT SYSTEMS 3/4.611 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY

LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

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

ACTION:

Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY

$b within9ane hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following hours.

a SURVEILLANCE REQUIREMENTS i

4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a. At least once per 31 days by ver ifying that:

All penetrations m not capable of being closed by

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1.

OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except for valves that are open under administrative control as permitted by Specification 3.6.3.1.

2. All equipment hatches are closed and sealed.

b. By verifying that each containment air lock is in

compliance with the requirements of Specification 3.6.1.3.

(c . Af t-r eachexcept cleeingth0 of cent:irrent each pe--tratia=]e.uhj-ct l!!tc Ty"- n) tecting, ir le:M , Opened feliceing : Type A er B tect, by leth rate tecting the 22:1 with gee et preeeure net leer thir P. ('.0. 0 prig) , :nd hMk "^rifying that "h=" th^ recrured le Mage rite for there cetle le dded to the leekage retee deterrined.pureuant to Sp ific: tion 4.0.1.2.d for all other Type 0 and C penetratione, the cc=hined lecheg: rat; i: 1;;; than 0 . 0 0 L. .

(1) Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed, or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days.

BEAVER VALLEY - UNIT 1 3/4 6-1 Amendment No. 499 (Fr.peje) Wodh

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- DPR-66 I'N' N '" O # W'b CONTAINMENT SYSTEMS fpec,$,ce,htn 6 eel 4dItd CONTAINMENT LEAKAC'g; ** *1 * **

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} LIMITING CONDITION FOR OPERATION l

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3.6.1.2 containment leakage rates shall be .ir.it.d i tw.

. . ' - Overell i=te; ret ' 1--' ; ::t: Of; l 1

) 1. : !.. , 0.10 per:: t 5.* rei;ht Of th: :;nt inz.e..t 01:

OREE 2: e .;:: n h: . : et e., <30.e :i;;, er i

b. ' :: dined le M ;e rete ef d 0.50 '. fer 211 ;:::t ti:n: ;

i :nd T:1r : e d iret te */;c_?_-.' c "--tm v'.:n ;r_;;uri =? l

! t0 ".f Mi\% 4he. cenWo mewh Ie.tl e.h3e i -h e. Imdr 3 Perleet 4w. the e.ba !

APPLICABILITY: MODES 1, 2, 3 and 4. }4. w&,4 to,[3 ui,W,n i bourer bein l ACTION: M lud HeT57ANOGYwhhors o^& m cot.

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l, SURVEILLANCE REQUIREMENTS l

! 4.6.1.2 The containment leakage rates shall be . ::nct ;t;d :t th:

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BEAVER VALLEY - UNIT 1 3/4 6-2 Amendment No.485--

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[Proysteb Y

PR-66 f __

NTAINMENT SYSTEMS SU EILLANCE REQUIREMENTS (Continued) l

b. If any periodic Type A test fails to meet .75 L , the test schedule for subsequent Type A tests shall be r viewed l nd approved by the Commission. If two consecutiv Type A l t sts fail to meet .75 L, a Type A test hall be l pe formed at least every $8 months until two nsecutive Typ A tests meet .75 L a at which time th above test sche le may be resumed.

c. The a uracy of each Type A test shall b verified by a suppleme tal test which:

1. Confi s the accuracy of the Type test by verifying that e difference between sup lemental and Type A test da is within 0.25 L a*

2. Has a du ation sufficient to curately establish the change in leakage for betwee the Type A test and the supplemental test.

3. Requires the quantity f gas injected into the containment or bled fr the containment during the l supplemental tes to equivalent to at least 25 percent of the otal measured leakage rate at P a (40.0 psig).

d. Type B and C tests 11 be conducted with gas at Pa *(40.0 psig) at in rva s no greater than 24 months except for tests invol ing:

1. Air locks,

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2. , Penetrations using conti ous leakage monitoring systems, an

3. Valves pr ssurized with fluid fr.m a seal system.

e. Air locks hall be tested and demo trated OPERABLE per Surveillan Requirement 4.6.1.3.

f. Leakage from isolation valves that are ealed with fluid from seal system may be excluded, ubject to the provi ons of Appendix J,Section III.C.3, w en determining the ombined leakage rate provided the s 1 system and val es are pressurized to at least 1.10 P a (44.0 psig) an the seal system capacity is adequate to main ain system p essure for at least 30 days.

A licable valves may be tested using water as the pr sure luid in accordance with the Inservice Testing Program. '

VER VALLEY - UNIT 1 3/4 6-3 Amendment No. 44-3--

<cm, ,.a e>,as,3

. DPR-66 '

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

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BEAVER VALLEY - UNIT 1 3/4 6-4 h

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1 DPR-66 l CONTAINMENT SYSTEMS l CONTAINMENT AIR LOCKS l

LIMITING CONDITION FOR OPERATION l 3.6.1.3 Two containment air locks shall be OPERABLE: j l

APPLICABILITY: MODES 1, 2, 3 and 4.

1 ACTION:

- - - - - - - - - - - - - GENERAL NOTES - - - - - - - - - - - - - -

1. Entry and exit is permissible to perform repairs on the affected air lock components.

2. Separate ACTION statement entry is_ allowed for each air lock.

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3. Enter the ACTION of LCO 3. 6.1.1h when air lock leakage resultsinexceedingthecombinedcontainmentleakagerate$ 6tccp % <e.crsh & ,

a. With one containment air lock door inoperable in one or more containment air locks:"'

1. Verify the OPERABLE door is closed in the affected air lock within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and

2. Lock the OPERABLE' door closed in the affected air lock within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and

3. Verify the OPERABLE door is locked closed in the af fected air lock at least once per 31 days. "'

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4. Otherwise, be in at least HOT STANDBY within the next I 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 )

hours. '

. l (4) Entry and exit is permissible for 7 days under administrative controls to perform activities not related to the repair of affected air lock components.

(5) Air lock doors in high radiation areas may be verified locked closed by administrative means. I 1

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, BEAVER VALLEY - UNIT 1 3/4 6-5 Amendment No. 4W- !

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l 4. B y PRtwiynk.}.A, k 46, DPR-66 a,e6fgd W

< CONTAINMENT SYSTEMS g g ,, y Le. e. Ted g Frpm *

SURVEILLANCE REQUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE

Sithir 72 ' cur:# fell ^':!ing Orth centeir cnt :ntry,

pt "her the air lech ir being ured fer multipic cntrie , then :t 100:t :nce p;r 72 h;urs,* by I

Ned '/ rifying no detectable seal leakage when the gap between the door seals is pressurized for at least 2

minutes to: {>,,p.(400gsg),

f)f. Personnel air lock grc:ter thir er equal t an n _s-

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_ l0,0 psy .

4 b) . Emergency air lock ~ renter thir _r equ:1 to

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y 10 0 F 19- undif h or, by qu:ntifying* the air lock q door seal leakage

. to ensure that the leakage rate is .es; than er equal d m .It3 ~ 4Mr 0 . 0 0 0 5 L., a t4 P. (40.0 psig) for the personnel air lock and ler: th:r er equ:1 *^ 0005 L a 10.0 psig j for the emergency air lock.

S* *

t. By r.ductina overal-1 air lock leakage tests ' at

n: t 12: thEr P. (40.0 psig), and verif

. overall air lock leakage rate is ' :: th:r er equal r j (40.0 psig):

te. 0. 05 L tat

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, At le;;t OnC; per 0 20nthe,# and

f 2. Prior ts ;;t;bli hing COMT.^.!P'EMT INTECP.ITY "h^n

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N b t dt*j 9 ::inten:nce h:h beer perferred er th: cir leek 10;k S r *t d d d m O *- that : uld Offect th cir ;;;1ing

F:hility- L:::1 1;;h : t: t :tini t 2 c e d e e d L t d .Nt. -precepre Of :t 1::: th:n P ::y be rub:tituted g 74*) p,O,.,I ad 4

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DELETE-y

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The previsiens of Opecification 4.0.2 are not epplicabl h7)

(h An inoperable air lock door does not invalidate the previous successful performance of the overall air lock eakage test.

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Results shall be evaluated against the acceptance criteria ed LCO 3.6.1.2.

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F#5;rs ;E d,o,i_+i?i.f.f..u 3 ;0 10 c a 00, DELETE J BEAVER VALLEY - UNIT 1 3/4 6-Sb Amendment No. M (bro 7eeckMo

DPR-66.

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) i X. Following maintenance performed on the outer ,

, personnel air lock door which may result in a 7 decrease in closure force on any part of the door sealing surface.

b. )G At least once per 18 months during shutdown by verif ying%a ..q) i gg_ d. Da one door in each air lock can be opened at a time.

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BEAVER VALLEY'- UNIT 1 3/4 6-Sc Amendment No. M (htops5cb Wotbss

i DPR-66 !

\ - l 4 CONTAINMENT SYSTEMS CONTAINMENT STRUCTURAL INTEGRITX LIMITING CONDITION FOR OPERATION l 3.6.1.6 The structural integrity of the containment shall be maintained at a level consistent with the acceptance criteria in !

Specification 4.6.1.6.1.

APPLICABILITY: MODES 1, 2, 3 and 4.

) ACTION:

With the structural integrity of the containment not conforming to the above requirements, restore the structural integrity to within the limits ="ler te increaeing the Re: :ter C001:nt Oy;t;; t;;perature

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(dkm a we or 6 m 4 l..tk heT d'MND8T M% AM Md l j

@ bour3 o^d m COLO J}hrT00wW Mksn % My g %g, j SURVEILLANCE REQUIREMENTS 2

4.6.1.6.1 Containment vessel Surfaces The structural integrity of '

the exposed accessible interior and exterior surfaces of the

containment vessel, including the liner plate, shall be determined i >during the shutden fer each *ype S. centei- r.t 10th g: ::te t :t l (reference S; cificatien 4.5.1.2) by a visual ine; rt:.0- Of th:::

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i cent:irent leck;;; ret; t;;t te verify ne epper ni. G ouy.e in "a"-trinc^

<< ^" ^th-" th"^"'"al d"^,*!diti^".

4.6.1.6.2 Reports b initi:1 : ;0rt Of :ny hn =:1-degrediti Of the centainre".t structure detect ^d duri=g the chave r~;uired teste and in:pecti:n: 0h:11 bc ::d: Within 10 d.y: Ofter ;;;pleti;n of the

rveillen;; res;ir;;;nt; cf thi; ep;;ifi;;tien, ;nd th; d;t 44ed

! -r:p::t cheil bc : b=itt:d pur:::nt to Op;;ification 0.';.: within ^0 4

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gw w camm.s t.xp.h g erp BEAVER VALLEY - UNIT 1 3/4 6-10 Amendment No. &

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DPR-66 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 i 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 limits of 10 CFR 100 during accident conditions.

RWLAC.GuMTRil 3/4.6.1.2 CONTAINMENT LEAKAGE 7 q Q lt 4 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.. fAr 2r added

^:en:Orvatten, nn: ::: ur d ufer:11 integratec :::nrg :::: 10 further

/ limited t: ; 0.75 L. during perfernane ef the periedic teet to J ::: cunt for p:: ible degrad: tion of th: cent inment lechage barriers between 10 hage texte.

The curveillene t : ting fer me: Curing 10 hag rate: Or cenci tent with the requirc:ent ef Appendix "J" cf 10 CTR 50. .

/The exemptier te 10 cFP 50 App =adiv T 777 n '(=) =11au- myp= A tests to be ecnducted en : 40 1 10-month echedule, not in conjunction Uith

.- . 2. . .

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DELETE 2 3/4.6.1.3 CONTAINMENT AIR LOCKS BACKGROUND containment air locks form part of the containment pressure boundary and provide a means for personnel access during all MODES of operation.

Each air lock is nominally a right circular cylinder, with a door at each end. The doors are interlocked to prevent simultaneous opening.

During periods when containment is not required to be OPERABLE, the door interlock meenanism may be disabled, allowing both. doors of an air lock to remain open for extended periods when frequent containment entry is necessary. The emergency air lock, which is located in the equipment hatch opening, is normally removed from the containment building and stored during a refueling outage. Each air lock door hat, been designed and tested to certify its ability to withstand a pressure in excess of the maximum expected pressure following a Design Basis Accident (DBA) in containment. As i

BEAVER VALLEY - UNIT 1 B 3/4 6-1 Amendment No. N I l

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Attachment to 3/4.6.1.2 Containment Leakaag INSERT "A" Containment leakage is limited to 5 1. 0 La, except prior to the first startup after performing a required Containment Leakage Rate Testing Program ' leakage test. At this time additional leakage limits must be met. As left leakage prior to the first startup after performing a required leakage test is required to be

< 0.60 La on a maximum pathway leakage rate (MXPLR) basis for combined Type B and C leakage following an outage or shutdown that included Type B and C testing only and < 0.75 La for overall Type A leakage following an outage or shutdown that included Type A l testing. At all other times between required leakage rate tests, i the acceptance criteria is based on an overall Type A leakage limit I of i 1.0 La and < 0.60 La on a minimum pathway leakage rate (MNPLR) I basis. The MXPLR for combined Type B and C leakage is the measured leakage through the worst of the two isolation valves, unless a penetration is isolated by use of a valve (s), blind flange (s) , or de-activated automatic valve (s) . In this case, the MXPLR of the 4

isolated penetration is assumed to be the measured leakage through the isolation device (s).

i BEAVER VALLEY - UNIT 1 Proposed Wording

. DPR-66 CONTAINMENT SYSTEMS BASES 3/4.6.1.3 ' CONTAINMENT AIR LOCKS (Continued)

BACKGROUND (Continued) such, closure of a single door supports containment OPERABILITY.

Each of the doors contains double o-ring seals and local leakage rate testing capability to ensure pressure integrity. DBA conditions which increase containment pressure will result in increased saaling forces on the personnel air lock inner door and both doors on the emergency air lock. The outer door on the personnel air lock is periodically tested in a manner where the containment DBA pressure is attempting to overcome the door sealing forces.

The containment air locks form part of the containment pressure boundary. As such, air lock integrity and leak tightness is essential for maintaining the containment leakage rate within limits in the event of a DBA. Not maintaining air lock integrity or leak tightness may result in a leakage rate in excess of that assumed in the unit safety analyses. f0R 4.0.1.2 lesk;g; rat requircrents' Fcs; ply w:, n _ l e = te, 4.ppendix 2, se =^dified by ppreved; te @ , m f DELETE # -

APPLICABLE SAFETY ANALYSES The DBAs that result in a release of radioactive material within containment are a loss of coolant accident and a rod ejection accident. In the analysis of each of these accidents, it is assumed ,

that containment is OPERABLE such that release of fission products l

), is controlled by the rate of containment to the environment leakage. The containment was designed with an allowable leakage [

rate of 0.1 percent of containment air weight per day. This leakage i rate is defined in310 Orn 50, Apper. dix A as L. = 0.1 percent of containment air weight per day, the maximum allowable containment leakage rate at the calculated peak containment internal pressure '

4 P. = 40.0 psig following a DBA. This allowable leakage rate forms

the basis for the acceptance criteria imposed on the SRs associated with the air locks.

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BEAVER VALLEY - UNIT 1 B 3/4 6-2 Amendment No. O

DPR-66 CONTAINMENT SYSTEMS l BASES 3/4.6.1.3 CONTAINMENT AIR LOCKS (Continued)

ACTIONS (Continued).

in which the OPERABLE door is expected to be open. At no time should the OPERABLE door be opened if it cannot be demonstrated that '

the inoperable door is sufficiently closed / latched. This verification is necessary to preclude an inadvertent opening of the

inoperable door while the OPERABLE door is open. After each entry and exit, the OPERABLE door must be immediately closed. If ALARA conditions permit and personnel safety can be assured, entry and exit should be via an OPERABLE air lock.

General Note (2) has been added to provide clarification that, for this LCO, separategction statement entry is allowed for each air lock. {geygp}j ogtp ec_ egeg In the event the air lock leakage results in exceeding the combined containment leakage rate General Note (3) directs entry into the gcquired ?.cti:n of LCO 3.6.1.1 ] go3434

a. With one air lock door in one or more containment air locks 1 the OPERABLE door must be verified closed TMN sMewh 4(..: quired operable, *:tica lock.

a.1) in each affected containment air This ensures that a leak tight containment barrier is maintained by the use of an OPERABLE air lock door.

This action must be completed within 1 hour. This specified time period is consistent with the ..cquired

?.ctic. of LCO 3.6.1.1 which requires CONTAINMENT INTEGRITY 11 la.

to be restored within~1 hour. g go3,y,g a m

_In addition, the affected air lock penetration must be isolated by locking closed (N:quircd 7.ctica a.2) the OPERABLE air lock door within the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months gompletion gime.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months /ompletion gime is reasonable for locking the OPERABLE air lock door, considering the OPERABLE door of the af fected air lock is being maintained closed. This action places additional positive controls on the use of the air lock when one air lock door is inoperable.

k.aquired ?.ction a has been modified by a Note.

Note (4) allows use of the air lock for entry and exit for 7 days under administrative controls. Containment entry may be required to perform non-routine Technical Specification '

(TS) Surveillances and Required Actions, as well as other activities on equipment inside containment that are required by TS or activities on equipment that support TS-required equipment. An example of such an activity would be the isolation of a containment penetration by at least one operable valve, and the subsequent repair and BEAVER VALLEY - UNIT 1 B 3/4 6-4 Amendment No. N lProps3*b Woe

s l

, DPR-66 l CONTAINMENT SYSTEMS i BASES 3/4.6.1.3 CONTAINMENT AIR LOCKS (Continued)

ACTIONS (Continued) post-maintenance technical specification surveillance testing on the inoperable valve. In addition, containment entry may be required to perform repairs on vital plant equipment which, if not repaired, could lead to a plant transient or reactor trip. This Note is not intended to preclude performing other activities (i.e., non-TS-required activities or repair of non-vital plant equipment) if the containment is entered, using the inoperable air lock, to perform an allowed activity listed above. This allowance is acceptable due to the low probability of an event that could pressurize the containment during the short time that the OPERABLE door is expected to be open.

gy 3bb 90 cquired ^ctier a.3 verifies that an air lock with an inoperable door has been isolated by the use of a locked and closed OPERABLE air lock door. This ensures that an acceptable containment leakage boundary is maintained. The i g'ompletion fime of once per 31 days is based on engineering judgment and is considered adequate in view of the low likelihood of a locked door being mispositioned. ncquired

""etion a.3 is modified by a Note (5) that applies to air lock doors located in high radiation areas and allows these l doors to be verified locked closed by use of administrative '

means. Allowing verification by administrative means is

' considered acceptable, since access to these areas is typically restricted. Therefore, it is unlikely that a door would become misaligned once it has been verified to be in the proper position.

b. With an air lock interlock mechanism inoperable in one or j more air locks, the AR0guired ?.cticn and associated )

dompletion dimos are consistent with those specified in )

00guired ?.ctigr a. ggQ sWM r ] Theb:quir:dAction: have been modified by two Notes. Note l (6) allows entry into and exit from containment under the

[a a control of a dedicated individual stationed at the air lock i to ensure that only one door is opened at a time (i.e., the individual performs the function of the interlock). Note I (5) applies to air lock doors located in high radiation 1 areas and allows these doors to be verified locked closed by use of administrative means. Allowing verification by ;

administrative means is considered acceptable, since access l

l l

Amendment No. #

BEAVER VALLEY - UNIT 1 B 3/4 6-5 )

(fr. pose) Wo )

l l '

l l 1

, DPR-66 )

l CONTAItiMENT SYSTEMS BASES 3/4.6.1.3 CONTAINMENT AIR LOCKS (Continued) j

.h j ACTIONS (Continued) l l

to these areas is typically restricted. Therefore, it is - I unlikely that a door would become misaligned once it has l been verified to be in the proper position. cnag acmON M Med) N

g. With one or more air than those described ing_cquired xs inoperable for reasoA

.'. ti n othery a or b,tnequired

'e, , beh eckM Ectier c.1 requires action to be initiated immediately to ,

d evaluate previous combined leakage rates using current air doors shaperMt. lock test results. An evaluation is acceptable, since it y 4gog is overly conservative to immediately declare the containment inoperable if both doors in an air lock have M *'h "^ ^ failed a seal test or if the overall air lock leakage is 1%penu *# I not within limits. In many instances (e.g., only one seal 4 g los4 per door has failed), containment remains OPERABLE, yet only 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (per LCO 3.6.1.f) would be provided to restore chood LWof*"g~J the air lock door to OPERABLE status prior to requiring a

' plant shutdown. In addition, even with both doors failing '

the seal test, the combined containment leakage rate can still be within limits.

@d W 6.W Required ^_ction c.2 requires that one door in the affected

~

T/containmentairlockmustbeverifiedtobeclosedwithin the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months dompletion consistent with the % gime.

This specified time period is quired ^.cticne of LCo 3. 6.1.1* which_

MTMN sW3 requires that CONTAINMENT INTEGRITY be restored within l' .

u (\ h 3.6.\id d Additionally, nagired J.ction c.3 requires that the must be restored to OPERABLE status

! af fected within the air24 lock (s)gompletion gime. The specified time hour period is considered reasonable for restoring an inoperable air lock to OPERABLE status, assuming that at least one g gQs,w or is maintained closed in each affected air lock.

For all Rcquired ^ctiene, if the inoperable containment air lock cannot be restored to OPERABLE status within the required gompletion dime, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The allowed gompletion gimes are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

BEAVER VALLEY - UNIT 1 B 3/4 6-6 Amendment No. M (ftt>poJed W0 l

NMENT SYSTEMS IDb b8 fy" Ted PN co m .

g BASES i 3/4.6.1.3 CONTAINMENT AIR LOCKS (Continued) Oh N N*ma I

Ledle Mt. EAg [togeg ' ,

SURVEILLANCE REOUIREMENTS (SRl l Maintaining containment air locks OPERABLE requires compliance with .

l the leakage rate test requirements of 10 CFR 50, Appendix J, e5 V l mcdified by--apprcved exemptiemk This SR reflects the leakage rate testing requirements with regard to air lock leakage (Type B leakage tests). The periodic testing requirements verify that the air lock leakage dcas not exceed the allowed fraction of the overall containment leakage rate. The frequency is tequired by ^ppendix-J, .

ac ,cdified--by c0creved OXemptione d Note (10) reflects the current?

Tappr-oved exemptier frer Appendix-2. Thuc, SP ' .0.2 (Ph ich-a4-lows '

OElf7&- N r09uon*Y-extencione) doe: not :PPlY : Stated ir "Ot; (?)=- j Testing of the personnel air lock door seals may be accomplished with the t.ir lock pressure equalized with containment or with atmospheric pressure. Each configuration applies P., as a minimum, across the sealing surfaces demonstrating the ability to function as designed. As long as the testing conducted is equivalent or more conservative than what might exist for accident conditions, the air lock doors will be able to perform their design function.

Performance of maintenance activities which affect air lock sealing capability, such as the replacement of the o-ring door seals and/or breach ring travel adjustment, will require performance of theQ appropriate surveillance requirements such as SR 4. 6.1. 3. af as a~

~ minimum. The performance of SR 4.6.1.3M will depend on the air h lock components which are affected by the maintenance.

of o-rings and/or breech ring travel adjustment on the Replacement inner personnel air lock door, for example, normally will not require the performance of SR 4.6.1.3.% as a post maintenance test. Testing per N SR 4.6.1.3.at is sufficient to demonstrate post accident leak tightness of the inner air lock door. The sealing force, which is applied to o-rings, is developed by the rotation of tapered wedges against the door's outer surface. This action forces the door to compress the o-rings which are located on the air lock barrel. When SR 4 . 6.1. 3. a+ is performed, the area between the two concentric o-rings is pressurized to at least P, and a leak rate of the two o-rings and sealing surface is determined. This test pressure applies an opposing force to the breech ring closure force. Since the containment pressure developed during a DBA applies a closing force which is supplemental to the breech ring force, the net result would be to improve the door sealing capability of the inner personnel air lock door over that which exists during the performance of SR 4.6.1.3.a. For this reason, performance of SR l.

BEAVER VALLEY - UNIT 1 B 3/4 6-7 Amendment No.4%

(frapow b Ust

. -DPR-66 j CONTAINMENT SYSTEMS 1 .

a y BASES

3/4.6.1.3 CONTAINMENT AIR LOCKS (Continued) f SURVEILLANCE REOUIREMENTS (SR) (Continued) 4 . 6 .1. 3 . 4s , ' which applies a force which opposes the ' breech ring force, is not necessary _.following certain inner air lock door maintenance. SR 4.6.1.3.atsufficiently demonstrates the' ability of j the inner air lock door to provide a leak tight barrier following l- -maintenance affecting the door sealing surface.

1 l Replacement of the o-rings on the outer personnel air lock door.,

which results in decreasing the breech ring closure- force, will a require performance of SR 4 . 6 .1. 3 .1r in addition to SR 4.6.1.3.a l which is required due to the door being opened. This surveillance 3

is required because containment DBA pressure tends to overcome the outer personnel air lock door sealing forces.- Performance of SR 4.

7 4.6.1.3.aton the outer personnel air lock applies an opposing force to the breech-ring closure force.in the same. manner as previously i described for the inner personnel air lock door. However,.for the

outer personnel air lock door, the containment pressure developed i during a DBA applies an opening force which is opposing the breech ring closure force. Therefore, upon completion of certain maintenance activities, continued outer door leak tightness during a l DBA cannot be assured by performance of SR 4. 6.1. 3.a4 alone. h-

Maintenance which may result in a decrease in closure force on any

! part of the door sealing surface, (decreasing of breech ring travel j- for example), will require performance of SR 4 . 6 .1. 3 .W The l performance of this surveillance is necessary to ensure that containment DBA pressure applied. against the outer door Will not

! result in the unseating of the air lock door by overcoming of the breech ring closure forces to the point where the leakage becomes h excessive. Since SR 4.6.1.3.4 duplicates DBA forces on the outer personnel air lock door and also measures the air lock leakage rate, performance of this surveillance requirement demonstrates the l continued ability of the outer personnel air lock door to provide a l leak tight barrier, during a DBA, following specific maintenance activities.

l The air lock interlock is designed to prevent simultaneous opening i of both doors in a single air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident containment pressure, closure of either door will support containment OPERABILITY. Thus, the door interlock feature 4 supports containment OPERABILITY and personnel safety, considering i the subatmospheric design, while the air lock is being used for

personnel transit in and out of the containment. Periodic testing of

this -interlock demonstrates that the interlock will function as '

designed and that simultaneous opening of the inner and outer doors will not inadvertently occur.

BEAVER VALLEY - UNIT 1 B 3/4 6-8 Amendment No.+99 W peMWo

.. . . . - . . . - _ . .. ~ . -. - . . - - _ - -.~

4

=.

- DPR-66

, CONTAINMENT SYSTEMS i

BASES q l

. 3/4.6.1.3 CONTAINMENT AIR LOCKS (Continued) i' g

SURVEILLANCE REOUIREMENTS (SR) (Continued) -

l- The SR has been modified by two h iti=:1) Notes. Note (7) states-that an inoperable air lock door coes not invalidate the previous successful performance of the overall air lock leakage test. This

is considered reasonable'since either air lock door is capable of

providing a fission product barrier in the event of a DBA. Note ( )

has been added to this SR requiring the results to be evaluated

against the acceptance criteria e4.LCO 3.6.1.2. ~

This ensures that i air lock leakage is properly accounted for' in determining the l combined containment leakage rate, gpheM,AW 3/4.6.1.4 and 3/4.6.1.5 INTERNAL PRESSURE AND AIR TEMPERATURE i The limitations on containment internal pressure and average air

' temperature as a function of river water temperature ensure that 1) the containment structure is prevented from ' exceeding its design I

~

negative pressure of 8.0 psia, 2) the containment peak pressure does

i. not exceed the design pressure of 45 psig during LOCA conditions, i and 3) the _ containment pressure is returned to' subatmospheric j- conditions following a LOCA.

The containment internal p.. essure and temperature limits shown as a i function of river water temperature describe the operational

envelope that will 1) limit the containment peak pressure to less than its design value of 45 psig and 2) ensure the containment i internal pressure returns subatmospheric within 60 minutes following

, a LOCA.

The limits on the parameters of Figure 3.6-1 are consistent with the

assumptions of the accident analyses.

1 3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY i This limitation ensures that the structural integrity of the j containment vessel will be maintained comparable to the original i design standards for the life of the facility. Structural integrity is required to ensure that the vessel will withstand the maximum pressure-of 40.0 psig in the event of a LOCA. The visual and Type A leakage tests gare sufficient to demonstrate this capability.

[) p erO tr A t.b d b 8 sg s'pteib.) tw Mt j gb.,4 Lat.g. 4 T.:by fpy

~

BEAVER VALLEY --UNIT 1 B 3/4 6-9 Amendment No. M

+(hofe5 %N

DPR-66

ADMINISTRATIVE CONTROLS l

-CORE OPERATING LIMITS REPORT (Continued)

4. T. M. Anderson to K. Kniel (Chief of Core Performance Branch, NRC) January 31, 1980 --

Attachment:

Operation and Safety Analysis Aspects of an Improved Load Follow Package.

Methodology applied for the following Specification: 3.2.1, Axial Flux Difference-Constant Axial Offset Control

5. NUREG-0800, Standard Review Plan, U. S. Nuclear Regulatory i Commission, .Section 4.3, Nuclear Design', July 1981. Branch Technical Position CPB 4.3-1, Westinghouse Constant Axial Offset Control (CAOC), Rev. 2, July 1981. Methodology applied for the following Specification: 3.2.1, Axial Flux Difference-Constant Axial Offset Control The core operating limits shall be determined so that all applicable I

limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic

! limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are l 4

met. The CORE OPERATING LIMITS REPORT, including any mid-cycle l revisions or supplements thereto, shall be provided on issuance, for each reload cycle, to the NRC Document Control Desk.

SPECIAL REPORTS

, 6.9.2 Special reports shall be submitted to the U. S. Nuclear Regulatory Commission, Document Control Desk, within the time period l specified for each report. These reports shall be submitted i covering the activities identified below pursuant to the requirements of the applicable reference specification:

a. ECCS Actuation, Specifications 3.5.2 and 3.5.3.

b. Inoperable Seismic Monitoring Instrumentation, Specification 3.3.3.3.

c. Inoperable Meteorological Monitoring Instrumentation, Specification 3.3.3.4.

d. Seismic event analysis, Specification 4.3.3.3.2.

. e. Sealed source leakage in excess of limits, Specification 4.7.9.1.3.

f. Miscellaneous reporting requirements specified in the Action Statements for Appendix C of the ODCM.

g. Contain;:nt In;;;; tion n;;;rt, Cp:;ification 4.0.1.0.2'.

O ELETED)

BEAVER VALLEY - UNIT 1 6-20 Amendment No.iab hooge) do

DPR-66 ADMINISTRATIVE CONTROLS

-OFFSITE DOSE CALCULATION MANUAL (ODCM) (Continued)

c. Shall be submitted to the Commission in the form of a j complete, legible ~ copy of the entire ODCM as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to i the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly

indicating- the area of the page that was changed, and shall indicate the date (e.g., month / year) the change was implemented.

Moved to the PROCESS CONTROL PROGRAM.

6.16

^

% T 9ss RT " B a

4 J

i i ,

l 1

i i

)

BEAVER VALLEY - UNIT 1 6-25 Amendment No. -14&

fogo3r)Wdm

- j INSERT "B" j 6.17 Containment Leakaae Rate Testina Procram A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 ,

CFR 50, Appendix J, Option B, as modified by approved I exemptions W . This program shall be in accordance with the )

guidelines contained in Regulatory Guide 1.163, " Performance-Based Containment Leak-Test Program," dated September 1995.

The peak calculated containment internal pressure for the design l basis loss of coolant accident, Pa, is 40.0 psig. j 1

The maximum allowable containment leakage rate, La, at P., shall be 0.10% of containment air weight per day.

Leakage Rate acceptance criteria are:

a. Containment leakage rate acceptance criterion is 5 1.0 La for the overall Type A leakage test and < 0.63 La for the Type B and Type C tests on a minimum pathway leakage rate (MNPLR) basis. During the first unit startup following testing in accordance with this program, the leakage rate acceptance < 0.60 La on criteria g e basis a maximum pathway )

leakage rate (MXPLR) for Type B and Type C tests ;

and < 0.75 La for Type A tests. 1

b. Air Lock testing acceptance criteria and required action are as stated in Specification 3.6.1.3 titled " Containment

]

Air Locks."

1 The provisions of Specification 4.0.2 do not apply to the test (

frequencies specified in the Containment Leakage Rate Testing i Program. l l

The provisions of Specification 4.0.3 are applicable to the !

Containment Leakage Rate Testing Program. l l

1 (1) Exemptions to Appendix J of 10 CFR 50 dated November 19, 1984, l December 5, 1984 and July 26, 1995. l (2) For penetrations which are isolated by use of a closed valve (s), blind flange (s) , or de-activated automatic valve (s) , l the MXPLR of the isolated penetration is assumed to be the l measured leakage through the isolation device (s). !

l 1

l l

l 1

l BEAVER VALLEY - UNIT 1 Proposed Wording l l

) '

ATTACHMENT A-2 Beaver Valley Power Station, Unit No. 2 l Proposed Technical Specification Change No. 94 I

i j

The following is a list of the affected pages:

I

! Affected Pages: XVI 3/4 6-1 ;

j 3/4 6-2 1 j 3/4 6-3 I i

3/4 6-4 I j 3/4 6-Sa 3 3/4 6-5b 3/4 6-9 B 3/4 6-1 i B 3/4 6-2 i B 3/4 6-4 B 3/4 6-5 B 3/4 6-6 .

B 3/4 6-7 j B 3/4 6-8 i B 3/4 6-9 I

6-21 i 6-25 I

i i

b 4

I l l

+

l l

l

. i

\

1 i

4

_ _ . _ . _ - - - _ + - - - - - _ _ . . - _ . . _ . - - . - - . - - - . - , - s

NPF

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE

, 6.7 SAFETY LIMIT VIOLATION . . . . . . . . . . . . . . .. 6-12 6.8 PROCEDURES . . ................... 6-12 6.9 REPORTING REOUIREMENTS 6.9.1 ROUTINE REPORTS . . . . . . . . . . . . . . 6-16 i 6.9.1.1 Startup Reports . . . . . . . . . . . . . . 6-16 6.9.1.4 Annual Reports. . . . . . . . . . . . . . . 6- 17 6.9.1.6 Monthly Operating Report. . . . . . . . . . 6-18 6.9.1.10 Annual Radiological Environmental Operating Report. .................. 6-19 ,

6.9.1.11 Annual Radioactive Effluent Release Report. 6-19 6.9.1.12 Core Operating Limits Report. . . . . . . . 6-19 6.9.2 SPECIAL REPORTS . . . . . . . . . . . . . . 6-20 i

6.10 RECORD RETENTION. . . . . . . . . . . . . . . . . 6-21 6.11 RADIATION PROTECTION PROGRAM. . . . . . . . . . . . 6-23 6.12 HIGH RADIATION AREA . . . . . . . . . . . . . . . 6-23 6.13 PROCESS CONTROL PROGRAM (PCP) . . . . . . . . . . . 6-24 6.14 OFFSITE DOSE CALCULATION HANUAL (ODCM). . . . . . . 6-25 ,

6.16 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT I SYSTEMS (Licuid. Gaseous and Solid) . . . . . . . . 6-25 l 6,,11 C ONTAINMEWT LE AKME 4 ATE TFrn kl6 PRosRAM . .. G-QS A0b 1 BEAVER VALLEY - UNIT 2 XVI Amendment No. 11r l ee) Worb )

a

'NPF-73 3/4.6 CONTAINMENT SYST EMS 3 / 4 . 6 '.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY

! LIMITING CONDITION FOR OPERATION i

3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

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

I

j. ACTION:

i

() Withou. primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

and in COLD SHUTDOWN within the following 46 hours5.324074e-4 days 0.0128 hours 7.60582e-5 weeks 1.7503e-5 months .

SURVEILLANCE REQUIREMENTS 30 j 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a. At least once per 31 days by verifying that

1

1. All penetrations

not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are 4

closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except i i for valves that are open under administrative control l as permitted by Specification 3.6.3.1. l

2. All equipment hatches are closed and sealed.

By verifying that each containment air lock is in

b. j compliance with the requirements of Specification 3.6.1.3. j l C. After ech cleeing Of each p?"etratier e"hject te Type k j t : ting, except the centairrent air lecke, if Opened '

, fellowing : Type A cr S tert, by leak rat ^ t-eting th= eeil with gae at e prese"re not lese +h== P. (24 ' peig), and

DEE eeritying that when the ce::urce le:x:ge ::te for th:::

i seele- 1: Odded te the 1 skege rater d=tareiaad p"r="="* *n

. spee:fic: tion i.6.,1.2.d fer :11 ether Type 9 ead C per.etrat iene , the cembi"-d leshige rate i: 1::: th:r

0 . 6 ^ L. . ;

i (1) Except valves, blind flanges, and deactivated automatic valves 4

which are located inside the containment and are locked, sealed, or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per

.92' days.

BEAVER VALLEY - UNIT 2 3/4 6-1 Amendment No.49-l {htspschLUst

_ - - _ _ - - _ _ _ . - - - . . - .- . . - - . -- - _ - . - _ . - -_ - ~ .

l .

i a i

)

f @,\d .M accedene. 'uJsh ' S et t tkt'410h NPF-73 j CONTAINMENT SYSTEMS

g,y) 4,wd "66.4 t. My Q' fg'{'*d f CONTAINMENT _ LEAKAGE 4 i pr yo m i

j LIMITING CONDITION FOR OPERATION 1

i

! y' ' "--

3.6.1.2 Containment leakage rates shall be T. '-

... 07- 211 4.d :;-et-' le=ht e r et e ^

I " ,, ^ . 1^ - ree.t by

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! APPLICAE*LITY:

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SURVEILLANCE REQUIREMENTS

! l l

I 4.6.1.2 The containment leakage rates shall be

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J j 4, 61.3.f l (Ne.sk P *d' 'I 31 l

BEAVER VALLH - UNIT 2 3/4 6-2 Amendment No:M*

ke(

l __

y l

. IPF-73

' ^' OBETE CO AINHENT SYSTEMS

\

SURVEI LANCE kdOUIREMENTS (Continued) / \

c. e accuracy of each Type A test shall be verified by a su lemental t t which:

1. onfirms the accuracy of the Type A test by verit ing that the

. fference between supplemental and Type A test ata is within O. 5 L,.

2. Has duration sufficient to accurately e-t lish the change in leaka rate between the Type A test and c supplemental test.

3. Requires the quantity of gas injected i o the containment or bled from he containment during the s plemental test to be equivalent o at least 25 percent of e total measured leakage rate at P, ( 4.7 psig).

. d. Type B and C tests all be conducted ith gas at P (44.7 psig) at intervals no greater t. an 24 months xcept for tests involving:

1. Air locks,

2. Penetrations using con in us leakage monitoring systems, and t

3. Valves pressurized wit uid from a seal system.

e. Air locks shall be teste and de nstrated OPERABLE per Surveillance

Requirement 4.6.1.3.

f. Leakage from isolatio valves that a e sealed with fluid from a seal system may be exclu d, subject to th Provisions of Appendix J,

Secticn III.C.3, w en determining the mbined leakage rate provided the seal system d valves are pressuri d to at least 1.10 P, (49.2 psig) an the seal system capacity 1 adequate to maintain system preau for at least 30 days.

g. All test 1 kage rates shall be calculated us g observed data con-verted to bsolute values. Error analyses shal be performed to determi the inaccuracy of the measured leakage ates due to maximum measur ent accuracy and instrument repeatability, the measured leaka e rates shall be adjusted to include the meas rement error.

Ap[icablevalvesmaybetestedusingwaterasthepressurefluidin a ordance with the Inservice Testing Program EAVER VALLEY - UNIT 2 3/4 6-3

. C9 toped were)g )__ s

NPF-73 CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Two containment air locks shall be OPERABLE:

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

- - - - - - - - - - - - - GENERAL NOTES - - - - - - - - - - - - - -

l

1. Entry and exit is permissible to perform repairs on the ;

affected air lock components.

2. Separate ACTION statement entry is allowed for each air lock. y 3,g,g,Q

3. Enter the ACTION of LCO 3.6.1.1 when air lock leakage

results in exceeding the combined containment _ leakage rate

@cephet craenca

a. With one containment air lock door inoperable in one or more containment air locks:"'

1. Verify the OPERABLE door is closed in the affected air lock within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and

2. Lock the OPERABLE door closed in the affected air lock within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and

3. Verify the OPERABLE door is locked closed in ths af fected air lock at least once per 31 days. "'

4. Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

(4) Entry and exit is permissible for 7 days under administrative controls to perform activities not related to the repair of affected air lock components.

(5) Air lock doors in high radiation areas may be verified locked closed by administrative means.

BEAVER VALLEY - UNIT 2 3/4 6-4 Amendment No. O

( d opco d W b M '

NPF-73 ca. 6'I pidetmad M.in%uhe8s air lac.

conTunMenT systems wa..y e.\a ierg.

5 p et & t h 4 M bd*M*O s %. km "3 L**N s.

3)N SURVEILLANCE REQUIREMENTS UN E ,

l b

1 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

. "ithin 72 h:ur " f:ll:ving ;;;h ;;nt:1. ::nt :nt ry.,

ext:;t '.cr th: cir 10:h i: b:ing ::d f ;1ti

ntrier, th r :t 1:::t sn;; p;r 72 h;;rc,* ple Ly

g,\)anf verifying no detectable seal hakage when the gap l between the door seals is pressurized for at least 2 minutes to: g (q j al g Personnel air lock ge::ter th an--c. :qu:1 to

"E N Q 1o.0 ysg ,

t Q p Emergency air lock

'V 'V n __z_

ter th:- :: :qu:1 to

} 7"*7 g i 4 dQ&W 2 or, b/ qu nti3y+nef') the air lock q door seal leakage to ensure that the leakage rate is _ :: th:r er equ:1 be 0.0005 L. 4 at

P.. (4 4. 7 -psig) for the personnel

3Q air lock and 1 :: then cr squ 1 t- 0 0005 L. at 10.0 j psig for the emergency air lock. 4 .

(a) g, (adM Ot. y : rducting overall air lock leakage tests at th:n P. (44.7 psig), and verifytikh the net ler

, overall air lock leakage rate is ' :: then :r q:.1 t

?

2 L. 44.7 psig): gg

'. t le::t :n:: p:: 0 ;; nth;," nd 4 ) M l b. I n, Ateau [.

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e aawn ww

_ _ u , 2 _ u 2 _ _. _ . . m, m,_

was w usw e a ty4 4 4 a sy u._ u___

%s wai o s sa sis aadai e

___,____2 __

,. i.. .m m ,. . ,

a si a a.sw asa e a su_ .,_

..t__

w a aiiis a a

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=

c 4ML,ub s. a L '. _ L_'

1'_'l.',

2_ "'.Z. .J_ __. _'_T ' " Z. .E

. . _ ' ".7.'_

U_. .g" M s Te g Pr.gr % ; ::; bility. L:::1 lech ::t; tc.; ting at

, o,A preceure Of net ler th:n " ::y b; zub:titut:d f r :n Over:ll cir 1 h t::t uh;r: the d::ign '

permite,": nd-DE LETE ---

v - - -

h7) _ "'h: pr;vici:n; cf Op::ific ti n 4.0.2 cr; n t r.pplicable.-

]

(1 An inoperable air lock door does not invalidate the 4

previous successful performance of the overall air lock _

! o goie...akage s cJ test.

(8 Results shall be evaluated against the acceptance criteria Oef LCO 3.6.1.2.

l(lk L._

Ex::; tier t 0;;r: tin; 1i

..;;:ndix rer.

J Of 10 CFPi 50, :: Ot:ted in the i del.ETE - T BEAVER VALLEY - UNIT 2 3/4 6-Sa Amendment No. &

(froyoseO N'y)

NPF-73

. CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

WX Following maintenance performed on the outer personnel air lock door which may result in a decrease in closure force on any part of the door sealing surface.

)(' At least once per 18 months. during shutdown by verifying QhA ,,gg OELE TE 4 onl)onedoorineachairlockcanbeopenedat a time.

I l l 1 l

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l 3

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BEAVER VALLEY - UNIT 2 3/4 6-5b Amendment No. O -

(fropseb wsrbml

l i NPF-73

- CONTAINMENT SYSTEMS i

j CONTAINMENT STRUCTURAL INTEGRITY i

i LIMITING CONDITION FOR OPERATION

! 3.6.1.6 The structural integrity of the containment shall be j maintained at a level consistent with the acceptance criteria in

Specification 4.6.1.6.1.

) APPLICABILITY: MODFS 1, 2, 3 and 4.

ACTION:

With the structural integrity of the containment not conforming to the above requirements, restore the structural integrity to within j the limits --ier_tc increeeiaa th= Reector C^^1:nt Sy ter ter; rature

--u-u---- ---sanew --

myn$yourer% cme,\le.dRoT$79)o h hw 6 ned3 l O*d **d sn Cot-D 3 bn)W W va hm .ht QQ '3 o bu,3 .J ~

SURVEILLANCE REQUIREMENTS

! 4.6.1.6.1 fd2Dtainment Vessel Surf aces The structural integrity of i the exposed accessible interior and exterior surfaces of the j containment vessel, including the liner plate, shall be determined j -9d uri ng the chutde"n fer each Type a cent =4a==a* 1==k=g= r=*= teet

- a t, w ._p ._ _w ._ n_ _ e_ m_ e n, n_ e_ _4 p t e_ n_ &_ _4 a_ A_

_ __ __ _ n_ a_ _ 1_ _ 9_ ,4 h_ o, _ u_ _t _ n_ _. _1 _4_=_

,_ =_ = _ &_ _4 a_ =_

_ a_ _s *k__._.._

eurfacee - This inepection eh=11 ' perfe- ed prier te the Type t centei= ent leekege rate teet te verify ne :pptrent ch:n; : 1:-

appearence er eth=r =hac--el d9 redation.

4.6.1.6.2 Renorts .f.n initi:1 :: pert Of :ny 05n::::1 de;r:dition of i

i the cent:i- ent structure detected during th: Obev: ::p ir:d t::t:

4-d inepectiene ehell ' rede ei6 kin 1^ day ef ter errpletier of the

curveillener: :pir: rent: Of thi: 0; ific tien, :nd the deteiled h- -"'=itt M pursuent te S;::ific: tion 5.0.:: ;;ithin 00

, > Fr* -h=ll ,

1 de"? ift-r 005pletien. Thi: repert chill includ: : d:::ripti:n Of th: : nditien Of the liner pl:te :nd cenerete, the ine; ction pr~-edure, the tolerene e en cracking :nd ""e cerrective ectiene i t ihe *. _

kes {<sguiq spc&d tw h Cm\ommtd LuW s. gRds" Teds. hr=* ) h a Uom\ swep&ie et be suchtti 1kV N b SkM " M W"' t ts we tosc\ewee o$ N eu d e \ b iorabten N "'dOO"*b k (co k,mw.A J 6 6 A 94 rh oc led %M**" r

[pepdaokce+bmwmVv@A Wptthw3 YhO he P f'f*b * **'C#'\*#"3 j g ws, cc Wn M lea =g. Ab Tobg Pr og r a x -

i i

BEAVER VALLEY - UNIT 2 3/4 6-9 Amendment No. 1 i

N Cf fof05 T

__ - - _ . ~ _ _ . . _ _ _ ___

NPF-73 3/4.6 CONTATNMENT 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 limits of 10 CFR 100 during accident 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 ass d in the accident analyses at the peak accident pressure, P.. A* aa dd:d cencer/:tice, the mercured c'/er:11 integr:ted 1,ccr:g: r;t: is further limited te d 0.75 L. during perfern:nt: Of the p riedic '

tect te acccunt fer percible degr:dati n Of th; : nt in;;nt 1; h:g:

barriere botu::n 10:h:g: t :tc.

The curveill:ne: tccting fer reaeuring leek =ga *=*a= =*=j Qcneirtent "ith the requirement -Of ?.ppendix "J" Of 10 CFR 50. j 3/4.6.1.3 CONTAINMENT AIR LOCKS _

ggQ h,)lT BACKGROUND l

containment air locks form p. art' of the containment pressure

. boundary and provide a means for personnel access during all MODES of operation. -

l l Each air lock is nominally a right circular-cylinder, with a door at each end. The doors are interlocked to prevent simultaneous opening. During periods when containment is not required to be OPERABLE, the door interlock mechanism may. be disabled,- allowing ~ both doors of an air lock to remain open. for extended periods when frequent containment entry is necessary. The emergency air lock, 3 which is located in the equipment hatch o:ening, is normally removed from the containment building and stored during a refueling outage.

Each air lock door . has been designed and tested to certify its ability to withstand a pressure in excess of'the maximum expected pressure following a Design Basis Accident (DBA) in containment. As

, such, closure of a single door supports containment OPERABILITY.

Each of the doors contains double o-ring seals and local leakage rate testing capability to ensure pressure integrity. DBA conditions which increase containment pressure will result in l

BEAVER VALLEY - UNIT 2 B 3/4 6-1 Amendment No. 4 (frop5eb urO

Attachment to 3/4.6.1.2 Containment Leakace INSERT "C" Containment leakage is limited to s 1.0 L., except prior to the first startup after performing a required Containment Leakage Rate Testing Program inakage test. At this time additional leakage limits must be met. As left leakage prior to the first startup after performing a required leakage test is required to be

< 0.60 L, on a maximum pathway leakage rate (MXPLR) basis for combined Type B and C leakage following an outage or shutdown that included Type B and C testing only and < 0.75 L, for overall Type A leakage following an outage or shutdown that included Type A testing. At all other times between required leakage rate tests, the acceptance criteria is based on an overall Type A leakage limit of i 1.0 La and < 0.60 L on a minimum pathway leakage rate (MNFLR) basis. The MXPLR for combined Type B and C leakage is the measured leakage through the worst o? the two isolation valves, unless a i penetration is isolated by use of a valve (s), blind flange (s) , or de-activated automatic valve (s) . In this case, the MXPLR of the isolated penetration is assumed to be the measured leakage through the isolation device (s).

l l

l I

l m

I f

BEAVER VALLEY - UNIT 2 Proposed Wording f

NPF-73 CONTAINMENT SYSTEMS BASES

3/4.6.1.3 CONTAINMENT AIR LOCKS (Continued) ,,

BACKGROUND (Continued) increased sealing forces on the personnel air lock inner door and both doors on the emergency air lock. The outer door on the personnel air lock is periodically tested in a manner where the containment DBA pressure is attempting to overcome the door sealing forces.

The containment air locks form part of the containment pressure boundary. As such, air lock integrity and leak tightness is essential for maintaining the containment leakage rate within limits in the event of a DBA. Not maintaining air lock integrity or leak tightness may result in a leakage rate in excess of that assumed in the unit safety analyses g SP ^ 6 1.2 leakage rate req"irerente, r_rply '!ith 10 CFP 56 Appendix 7, se modified by 2pprcved

'^

m- - - - - ,

OEl.ETE ]

APPLICABLE SAFETY ANALYSES 1 The DBAs that result in a release of radioactive material within containment are a loss of coolant accident and a rod ejection accident. In the analysis of each of these accidents, it is assumed that. . containment is OPERABLE such that release of fission products to the environment is controlled ~ by the rate of containment leakage. ~The containment was designed with an allowable leakage rate of 0.1 percent of containment air weight per day. This leakage

, rate is defined in .n10 CFP. 50, ^ppendix 2,. as L. = 0.1 percent of containment air weight.per day, the maximum allowable containment leakage. rate at the calculated peak containment internal pressure P. = 4.4.7 psig following a DBA. This allowable leakage rate forms the basis for the acceptance criteria imposed on the SRs associated with the air lock _s.

dnA6%(it?4M"(eM % d Le y M uhy k)ee

, LCD h.h coc.sainment air lock forms part of the containment pressure boundary. As part of containment, the air lock safety function is related to control of the containment leakage rate resulting from a DBA. Thus, each air lock's structural integrity and leak tightness are essential to the successful mitigation of such an event.

BEAVER VALLEY - UNIT 2 B 3/4 6-2 Amendment No.-M (Proposed W6M"

l

\

, NPF-73 CONTAINMENT SYSTEMS BASES 3/4.6.1.3 CONTAINMENT AIR LOCKS (Continued) i ACTIONS (Continued)

If ALARA conditions permit and personnel safety can be assured, entry and exit should be via an OPERABLE air lock.

General Note (2) has been added to provide clarification that, for this LCO, separate ^. tien statement entry is allowed for each

air lock.

c7f c, g g,,c, p n

In the event the air lock xleakage results in exceeding the

' combined containment leakage rat 6 General Note (3) directs entry into the 9:qu!r:d *. tica of LCO 3. 6.1.1Qawel 1.f.O '3. 6. l. Q

a. With one air lock door in one or more containment air locks inoperable, the OPERABLE door must be verified closed 9:quir d ?. tion a.1) in each affected containment air ock. This ensures that a leak tight containment barrier is maintained by the use of an OPERABLE air lock door.

This action must be completed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . This hg , ,specified time period is consistent with the Rcquired r^ction of LCO 3.6.1.1, which require)i CONTAINMENT INTEGRITY G L

"4; .

to be restored within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

GO bbbQ hjb"*

In addition, the affected air v lock penetration must be isolated by locking closed (= quired ?.ction a.2) the OPERABLEairlockdoorwithinthe24hourgompletiongime.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months gompletion gime is reasonable for locking the OPERABLE air lock door, considering the OPERABLE door of the affected air lock is being maintained closed. This action places additional positive controls on the use of the air lock when one air lock door is inoperable, ncquired ^.ction a has been modified by a Note. Note (4) allows use of the air lock for entry and exit for 7 days under administrative controls. Containment entry may be required to perform non-routine Technical Specification (TS) Surveillances and Required Actions, as well as other activities on equipment inside containment that are required by TS or activities on equipment that support TS-required equipment. An example of such an activity would be the isolation of a containment penetration by at least one operable valve, and the subsequent repair and post-maintenance technical specification surveillance testing on the inoperable valve. In addition, containment entry may be required to perform repairs on vital plant equipment which, if not repaired, could lead to a plant BEAVER VALLEY - UNIT 2 B 3/4 6-4 Amendment No. 4 ;

(Ag d Luod l

l

, NPF-73 -

l cot 1TAINMEllT SYSTEMS l BASES 3/4.6.1.3 CONTAINMEtIT AIR LOCKS (Continued) a ACTIONS (Continued) l transient or reactor trip. This Note is not intended to preclude performing other activities (i.e., non-TS-required activities or repair of non-vital plant equipment) if the containment is entered, using the inoperable air lock, to perform an allowed activity listed above. This allowance is acceptable due to the low probability of an event that could pressurize the containment during the short time that the OPERABLE door is expected to be open.

)nc qu i rc d ..cticr a.3 verifies that an air lock with an inoperable door has been isolated by the use of a locked hgd __ and closed OPERABLE air lock door. This ensures that an acceptable containment leakage boundary is maintained. The ,

5OW ! gompletion gime of once per 31 days is based on engineering judgment and is considered adequate in view of the low likelihood of a locked door being mispositioned. ncquired N Acti^= a.3 is modified by a Note (5) that applies to air lock doors located in high radiation areas and allows these doors to be verified locked closed by use of administrative 4 means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Therefore, it is unlikely that a door would become misaligned once it has been verified to be in the proper position.

b. With an air lock interlock mechanism inoperable in one or more air locks, the ?ncquired "ctienc and associated fompletion gimes are consistent with those specified in F- _

pgh _1 m__..im

.m---

,,,.4m.

---g a.

The equired ".ctienc have been modified by two Notes. Note

' SQMs" (6) allows entry into and exit from containment under the control of a dedic.ated individual stationed at the air lock to ensure that only one door is opened at a time (i.e., the individual performs the function of the interlock). Note (5) applies to air lock doors located in high radiation areas and allows these doors to be verified locked closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Therefore, it is unlikely that a door would become misaligned once it has been verified to be in the proper position.

BEAVER VALLEY - UNIT 2 B 3/4 6-5 Amendment Po.4 (keyo3e)

, NPF-73 CONTAIllMENT SYSTEMS l

l BASES 3/4.6.1.3 CONTAIUNENT AIR LOCKS (Continued)

@CT10N ske.h ACTIONS (Continued) g I

c. With one or more air locks y inoperable for reasonspther than those described in Reg ired 7. ti:n: a or b4n ,uired

?.cticr. col requires action to be initiated immediately to k,h, b.h oak ( evaluate previous combined leakage rates using current air lock test results. An evaluation is acceptable, since it

dmd M*

is overly conservative to immediately declare the '

a M m bclacK containment inoperable if both doors in an air lock have failed a seal test or if the overall air lock leakage is Nb d

not within limits. In many instances (e.g., only one seal m ,p m W o f per door has failed), containment remains OPERABLE, yet only 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (per LCO 3.6.1.1% would be provided to restore

\*N "# g the air lock door to OPERABLE status prior to requiring a aur3 mofrAhg- plant shutdown. In addition, even with both doors failing ;

the seal test, the combined containment leakage rate can still be within limits. Q uo 3,g, g,g "cquired ^.ctier c.2 requires that one door in the affected containment air lock must be verified to be closed within the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months gompletion time. This specified time period is F

consistent with the ";;uired ^.ctiene of LCO 3.6.1.1A which.

ACTjof\I / requires that CONTAINMENT INTEGRITY 3 be restored within 11 sW M \ " ""-

@P&W-""VJ 64 n *@!

'- Additionally, equir:d Acticr. c.3 requires that the!

must be restored to OPERABLE status af fected within air24 the lock (s)gompletion gime. The specified time ,

hour

, period is considered reasonable for restoring an inoperable Ug] air lock to OPERABLE status, assuming that at least one door is maintained closed in each affected air lock, i

$k%.%tN3 y k For all ncquired ?.:tions, if the inoperable containment air lock cannot be restored to OPERABLE status within the required d'ompletion gime, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at

.least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The allowed gompletion gimes are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

l i BEAVER VALLEY - bd 2 B 3/4 6-6 Amendment No. D Wespe&Gtc

--- - --- , . . - - _ - . - - - _ _ . - . . ~ . . . - . . - - . - . . - ~ - - - -

i i

! NPF-73 F 4 gI 4,, c 4..edt.. M g,y,,

CONTAINMENT SYSTEMS b%

l Te ht"^*

BASES 1

[, d . h Nhvetkl 3/4.6.1.3 CONTAINMENT AIR LOCKS (Continued)

{ A w r w ,- J T

SURVEILLANCE REOUIREMENTS (SR) l . .

{

MaintainingconkainmentairlocksOPERABLErequirescompliance i with the leakage rate test requirements f 10 0." 50, ?.;;:ndix J, 24 redified by :;;r:/ d x::;ti:nc. This SR reflects the leakage rate testing requirements with regard to air lock leakage (Type B leakage

, tests). The periodic testing requirements verify that the air lock leakaae does not exceed the allowed fraction of the overall I containment leakage rate. The grequency is*eg ired by A;;endix L

a: difi:d 5/ :::re/ d exerptiene J rete (le; rerie m * - m rr- *

, f_;;r: /:d extr; tion fr: ..;;:n :,x J .

Thu , On 4. 0. ;; (which alleW3

! m ETE 'r uciiue:27 extentic :) d: t 2FF1 7 :: etzted i- Met: (?) !

Testing of the personnel air lock door seals may be accomplished with the air lock pressure equalized with containment or with l atmospheric pressure. Each configuration applies P., as a 4

minimum, across the sealing surfaces demonstrating the ability to ,

function as designed. As long as the testing conducted is i j equivalent or more conservative than what might exist for accident ,

conditions, the air lock doors will be able to perform their design '

l function.

4

, Performance of maintenance activities which af fect air lock 4

sealingcapability,suchasthereplacementoftheo-ringdoorsealsM and/or breach ring travel adjustment, will require performance of the appropriate surveillance requirements such as SR 4. 6.1. 3.at as a minimum. The performance of SR 4.6.1.3%r will depend on the air

> d,Q lock components which are affected by the maintenance. Replacement

! of o-rings and/or breech ring travel adjustment on the inner j oorsonnel air lock door. for example,'normally will not require the i

nerformance of SR 4.6.1.3 3> as a post maintenance test. Testing per

SR 4.6.1.3.at is sufficient to demonstrate post accident leak l tightness 'of the inner air lock door. The sealing force, which is

eI applied to o-rings, is developed by the rotation of tapered wedges against the door's outer surface. This action forces the door to ;

. compress the o-rings which are located on the air lock barrel. When

! SR 4. 6.1. 3. at is performed, the area between the two concentric i o-rings is pressurized to at least P. and a leak rate of the two '

j o-rings and sealing surface is determined. This test pressure l i applies an opposing force to the breech ring closure force. Since

the containment pressure developed during a DBA applies a closing force which is supplemental to the breech ring force, the net result

- would be to improve'.the door sealing capability of the inner i

! personnel air lock door over that which exists during the

performance of SR 4.6.1.3.a. For this reason, performance of SR

. 1 i

BEAVER VALLEY - UNIT 2 B 3/4 6-7 Amendment No. *

(fropeeb %b

_ . _ _ _ - _ _ . _ _ _ _ _ - = . . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ - . - _ _ . .

11PF-73 CQNTAIMMENT SYSTEMS DASES 3/4.6.1.3 CONTAINMENT AIR LOCKS (Continued)

SURVEILLANCE REOUIREMENTS (SR) (Continued)

! 4.6.1.3. which applies a force which opposes the breech ring force, is not necessary following certain inner air lock door maintenance. SR 4.6.1.3.atsufficiently demonstrates the ability of ,

the inner air lock door to provide a leak tight barrier following ,

maintenance affecting the door sealing surface. l Replacement of the o-rings on the outer personnel air lock door, which results in decreasing the breech ring closure force, will .

4 a' require performance of SR 4.6.1.3 % in addition to SR 4. 6.1. 3.a which is required due to the door being opened. This surveillance ' i is required because containment DBA pressure tends to overcome the :

4 outer personnel air lock door sealing forces. Performance of SR l 4.6.1.3.a4on the outer personnel air lock applies an opposing force to the breech ring closure force in the same manner as previously described for the inner personnel air lock door. However, for the ;

outer personnel air lock door, the containment pressure developed during a DBA applies an opening force which is opposing the breech ring closure force. Therefore, upon completion of certain i

maintenance activities, continued outer door leak tightness during a 4 DBA cannot be assured by performance of SR 4.6.1.3.a+ alone.

i Maintenance which may result in a decrease in closure force on any part of the door sealing surface, (decreasing of breech ring travel {g, for example), will require performance of SR 4 . 6 .1. 3 . A. The

performance of this surveillance is necessary to ensure that e containment DBA pressure applied against the outer door will not

! result in the unseating of the air lock door by overcoming of the i breech ring closure forces to the point where the leakage becomes

' Q excessive. Since SR 4.6.1.3Nr duplicates DBA forces on the outer -

personnel air lock door and also measures the air lock leakage rate,

performance of this surveillance requirement demonstrates the continued a'bility of the outer personnel air lock door to provide a ,

4 leak tight barrier, during a DBA, following specific maintenance

, activities.

I The air lock interlock is designed to prevent simultaneous opening of both doors in a single air lock. Since both the inner and outer doors of an air lock are designed to withstand 'the maximum expected post accident containment pressure, closure of either door will support containment OPERABILITY. Thus, the door interlock feature supports containment OPERABILITY and personnel safety, i considering the subatmospheric design, while the air lock is being used for personnel transit in and out of the containment. Periodic '

testing of this interlock demonstrates that the interlock will

. ' function as designed and that simultaneous opening of the inner and l outer doors will not inadvertently occur.

i BEAVER VALLEY - UNIT 2 B 3/4 6-8 Amendment No.O l Sf%) !

NPF-73 CONTAIMMENT SYSTEMS BASES 3/4.6.1.3 CONTAINMENT AIR LOCKS (Continued)

SURVEILLANCE REOUIREMENTS (SR) (Contiaued) 1 .

The SR has been modified by two EMitic.7:3 Notes. Note ( )

states that an inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test. This is considered reasonable since either air lock door is

@ canable_,of providing a fission product barrier in the event of a DBA.

Note (N) has been added to this SR requiring the results to be evaluated against the acceptance criteria LCO 3.6.1.2. This ensures 'that air lock leakage is properly accounted for in determining the combined containment leakage rate. gaQ 3/4.6.1.4 and 3/4.6.1.5 INTERNAL PRESSURE AND AIR TEMPERATURE The limitations on containment internal pressure and average air temperature as a function of service water-temperature ensure,that

1) the containment structure is prevented from exceeding its de' sign negative pressure of 8.0 psia, 2) the containment peak pressure does not exceed the design pressure of 45 psig during LOCA conditions, and

3) the containment pressure-is. returned to subatmospheric conditions following a LOCA.

The containment internal pressure and temperature limits shown as a f unction .of service water temperature-describe the operational envelope that will 1) limit the containment peak pressure to less l than its design value of 45 psig and 2) ensure the containment .

)

internal pressure returns subatmospheric within 60 minutes following i a LOCA. Additional operating margin is provided if the containment i average air temperature is maintained above 100*F as shown on Figure 3.6-1.

The limits on the parameters of Figure 3.6-1 are consistent with the assumptions of the accident-analyses.

3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY-This limitation ensures that the . structural integrity of the containment 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 44.7 psig in the event of a LOCA. The visual and Type A leakagetests.gresufficienttodemonstratethiscapability.

i hekstN oN N het vt Spec h tb th Ot h m se d Leu bp ed R c. Tea bog r=$

BEAVER VALLEY - UNIT 2 B 3/4 6-9 Amendment No.

(lrspostb %bt

NPF-73

. ADMINISTRATIVE CONTROLS SPECIAL REPORTS (Continuedt

c. Inoperable Meteorological Monitoring Instrumentation, ,

Specification 3.3.3.4. '

d. Seismic event analysis, Specification 4.3.3.3.2.

e. Sealed source leakage in excess of limits, Specification 4.7.9.1.3.

f. Miscellaneous reporting requirements specified in the ACTION Statements for Appendix C of the ODCM.

g. Cent:inrent In:pertien ": pert, Sp::ific tien ? . C .1. 5. 2 9(EELEnE(D

h. Steam generator tube inservice inspection, Specification 4.4.5.5.

1. Inoperable accident monitoring, Specification 3.3.3.8.

j. Liquid Hold-Up Tanks, Specification 3.11.1.4.

k. cas Storage Tanks, specification 3.11.2.5.

1. Explosive Gas Monitoring Instrumentation, Specification 3.3.3.11.

6.10 RECORD RETENTION 6.10.1 The following records shall be retained for at least five (5) years;

a. Records and logs of facility operation covering time interval at each power level.

b. Records ed logs of principal maintenance activities, inspections, repair and replacement of principal items of i equipment related to nuclear safety.

c. All REPORTABLE EVENTS.

d. Records- of surveillance activities, inspections and calibrations required by these Technical Specifications.

e. Records of reactor tests and experiments.

'l BEAVER VALLEY - UNIT 2 6-21 Amendment No. 40-(frapyb Uhtb

NPF-73 ADMINISTRATIVE CONTROLS 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) t Changes to the ODCM: ,

a. Shall be documented and records of reviews performed shall be retained -as required by Specification 6.10.2.n. This documentation shall contain:

1) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and

2) A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, o r. setpoint calculations.

-b. Shall become effective after review and acceptance by the OSC and the approval of the General Manager Nuclear Operations, predesignated alternate or a predesignated Manager to whom t the General Manager Nuclear operations has assigned in writing the responsibility for review and approval of specific subjects.

c. Shall be submitted to the. Commission in the form of a complete, legible copy of the entire.0DCM as a part of or ,

with the Annual Radioactive Effluent Release concurrent Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly '

indicating the area of the page that was changed, and shall i indicate the date (e.g., month / year) the change was ,

implemented.

6.16 Moved to the PROCESS CONTROL PROGRAM.

< A00 T NSEU " O !

L t

)

i-

BEAVER VALLEY . UNIT 2 6-25 Amendment No. 74w Oropsch kNm

~ 1 j INSERT "D".

4

\ 6.17 CONTAINMENT LEAKAGE RATE TESTING PROGRAM

' leakage rate A program shall be established to implement the testing of the containment as required by 10 CFR 50.54 (o) and 10 Appendix J, Option B, as modified by approved J CFR 50, J

exemptions W . This program shall be in accordance with the guidelines contained in Regulat.ory Guide 1.163, " Performance-Based j containment Leak-Test Program," dated September 1995.

4 The peak calculated containment internal pressure for the design

- basis loss of coolant accident, Pa, is 44.7 psig.

l The maximum allowable containment leakage rate, L., at P., shall be j 0.10% of containment air weight per day.

Leakage Rate acceptance criteria aret

a. Containment leakage rate acceptance criterion is 1 1.0 La for the overall Type A leakage test and < 0.60 La for the

Type B and Type C tests on a minimum pathway leakage rate l (MNPLR) basis. During the first unit startup following testing in accordance with this program, the leakage rate 4

acceptance criteria g e basis < 0.60 La on a maximum pathway i leakage rate (MXPLR) for Type B and Type C tests and < 0.75 La for Type A tests,

{

b. Air lock testing acceptance criteria and required action 1 are as stated in Specification 3.6.1.3 titled " Containment

] Air Locks."

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

{ Program.

1 i The provisions of Specification 4.0.3 are applicable to the containment Leakage Rate Testing Program.

l-i i

i i - (1) Examptions to Appendix J of 10 CFR 50, as stated in the 3- operating license.

, (2) For penetrations which are isolated by use of a closed valve (s), blind flange (s) , or de-activated automatic valve (s) , l the MXPLR of the isolated penetration is assumed to be the .

. measured leakage through the isolation device (s). l i

i 1

BEAVER VALLEY - UNIT 2 Proposed Wording

-e., - .,- , -

n

ATTACHMENT B Beaver Valley Pcwor Station, Unit Nos. 1 and 2 Proposed Technical Specification Change No. 223 and 94 REVISION OF CONTAINMENT LEAKAGE TESTING REQUIREMENTS A. DESCRIPTION OF AMENDMENT REQUEST Limiting Condition For Operation (LCO) 3.6.1.1 titled

" Containment Integrity" would be revised. Specifically, the action statement would be revised by replacing the word "one" with the numerical value one. The words "following 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months " would be revised to read "following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ." In addition, Surveillance Requirement (SR) 4.6.1.1.c would be deleted.

LCO 3.6.1.2 titled " Containment Leakage" would be revised by replacing the specific numerical limits on containment leakage rates with a reference to Specification 6.17 titled

" Containment Leakage Rate Testing Program."

The Action statement of LCO 3.6.1.2 would be revised to require that the containment leakage rates be restored to within limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or the plant must be placed in cold shutdown within the following 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

SR 4.6.1.2 would be revised by removing the specific details on the required test intervals, test methodology and calculation of test results. Specifically, SR 4.6.1.2 would be revised by replacing the current reference to Appendix J of 10 CFR 50 with a now reference to the Containment Leakage Rate Testing Program. SR 4.6.1.2.a would be revised to require that Type A, B, and C testing, except for containment air lock testing, be conducted in accordance with the Containment Leakage Rate Testing Program. SR 4.6.1.2.b would be revised to reflect the wording similar to that contained in the current SR 4.6.1.2.e. For Beaver Valley Power Station (BVPS)

Unit No. 1 only, the current footnote designated by a single asterisk would be deleted. The remaining surveillance I requirements would be deleted.

The action statement of LCO 3.6.1.3 titled " Containment Air Locks" would be revised. Specifically, general note (3) would i be revised by adding the words "and 3.6.1.2,." In addition, l the words " acceptance criteria" would be added following the '

word " rate."

SR 4.6.1.3 would be revised by removing the specific details l on required test frequency. These specific details would be J replaced with wording which requires a test frequency as specified in the Containment Leakage Rate Testing Program.

The words "when tested" would be added following the terms 0.0005 La and 0.05 La. Footnotes number (7) and number (10) would be deleted. The remaining two footnotes would be renumbered to reflect the deletion of footnote (7). The current footnote (9) would be modified by adding the words l

ATTACHMENT B, continusd Proposed Technical Specification Change Nos. 223 and 94 i Page 2 s

" applicable to" following the word " criteria." In addition,

minor editorial changes would be made due to reformatting and l

replacing of words with symbols. The symbol for greater than

or equal to would be added to each specific test pressure not i previously denoted by this symbol to add consistency to this surveillance requirement.

i The action statement of LCO 3.6.1.6 titled " Containment l Structural Integrity" would be revised to require that

structural integrity be restored to within limits within i 1 hour or the plant must be placed in cold shutdown within the following 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

< SR 4.6.1.6.1 would be revised by removing the specific details on required test' frequency. The specific details on test

frequency would be replaced with wording which requires a test frequency as specified in the Containment Leakage Rate Testing Program. The acceptance criteria of the proposed SR 4.6.1.6.1 l would be revised to require that no evidence of structural 4 deterioration that might affect either the containment structural integrity or leak tightness be observed. SR 1 4.6.1.6.2 would be revised by removing the specific details on The specific details would be reporting requirements.

i replaced with wording which requires reports to be prepared in accordance with the Containment Leakage Rate Testing Program.

The Bases sections for 3/4.6.1.2, 3/4.6.1.3 and 3/4.6.1.6 would be revised to reflect changes made to each specification as previously noted. In addition, minor editorial changes would be made to add consistency between the Bases section and the Specifications and to provide additional clarification.

Specification 6.9.2.g of the Administrative Controls section would be deleted. Specification 6.17 titled " Containment Leakage Rate Testing Program" would be added to the i Administrative Controls Section. The Index would be revised to reflect the addition of Specification 6.17.

B. BACKGROUND i

The Nuclear Regulatory Commission (NRC) has amended its regulations to provide a performance based option, Option B, 4

for leakage rate testing of containments. Licenses may

voluntarily comply with this Option B as an alternative to compliance with the previous requirement of 10 CFR 50 Appendix J now known as Option A of Appendix J. Option B is aimed at improving the focus of the body of regulations by eliminating prescriptive requirements that are marginal to safety and by providing licensees greater flexibility for cost-effective implementation methods for regulatory safety

objectives. Option B of 10 CFR 50 Appendix J Section V.B titled " Implementation" requires that a request for revision to the technical specifications be submitted to the NRC which includes, by general reference, the regulatory guide or other B-2

l ATTACHMENT B, continu:d Proposed Technical Specification Change Nos. 223 and 94 Page 3 implementation document used to develop a performance-based-leakage testing program. This proposed amendment request contains the required reference to such document. In addition, a licensee must submit proposed technical specifications which would eliminate those technical specifications which implement the previous requirements contained in Option A. This proposed amendment request removes the prescriptive requirements of Option A concerning test frequencies and test methodology.

C. JUSTIFICATION The proposed revision to the action statement of LCO 3.6.1.1 to require the plant to be in cold shutdown within 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months will make this action consistent with the other LCOs pertaining to primary containment. The 30 hour3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months requirement is consistent with the NUREG 1431, Revision 1 titled " Standard Technical Specifications for Westinghouse Plants" (ISTS). The proposed revision of the word "one" to the numerical value one is also consistent with ISTS and other LCO action statements contained in the containment section of the technical specifications.

The proposed deletion of SR 4.6.1.1.c will remove the prescriptive details concerning retesting of penetrations subject to Type B testing. The prescriptive details on retesting of Type B penetrations are stated in Nuclear Energy Institute (NEI) 94-01, Revision 0, dated July 26, 1995,

" Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J." Specifically, section 10. 2.1. 3 of the NEI 94-01 provides guidance on retesting of Type B penetrations. The Containment Leakage Rate Testing Program requirements, contained in proposed Specification 6.17, states that this program shall be in accordance with the guidelines contained in Regulatory Guide (RG) 1.163 titled " Performance-Based Containment Leak-Test Program." RG 1.163 endorses the use of NEI 94-01. Therefore, the Containment Leakage Rate Testing program will continue to require retesting of Type B penetrations based on guidance which has been endorsed by the NRC. The proposed requirements for the Containment Leakage Rate Testing Program will continue to require that the measured leakage rate for ratested Type B penetrations, when combined with other Type B and C leakage rates, be less than 0.60 La.

The proposed amendment will remove the specific acceptance criteria for containment leakage rates from LCO 3.6.1.2. The specific details on containment leakage rate acceptance criteria are contained in proposed Specification 6.17. The proposed LCO 3.6.1.2 will directly reference Specification 6.17. For BVPS Unit No. 1 only, the limits specified in Specification 6.17 will allow the overall Type A leakage limit to be less than "or equal to" La. This change will make the B-3

, ATTACHMENT B, continutd

Proposed Technical Specification Change Nos. 223 and 94' Page 4

) Type A leakage limit consistent with current BVPS Unit No. 2 technical specifications and ISTS. Allowing the overall Type

! A leakage rate to be equal to La is consistent with the BVPS Unit No. 1 design basis leak rate as stated in the Updated Final Safety Analysis Report (UFSAR) Section 14.3.5.2. In addition, for BVPS Unit No. 1 only, the combined leakage rate will be further limited to less than (i.e., delete "or equal to") .60 L. which is consistent with the current BVPS Unit

} No. 2 technical specifications and ISTS.

The proposed revision to the action statement of LCO 3.6.1.2 is necessary to reflect that containment leakage rates can be i determined during plant operation (i.e. Modes 1 thru 4). As such, the appropriate action should be stated in this specification should the containment leakage rate limits be 4 exceeded during plant operation. The proposed action statement is consistent with the proposed action statement of LCO 3.6.1.1 and ISTS. Since compliance with LCO 3.6.1.1 I requires that containment leakage be within limits per the j definition of Containment Integrity, the proposed action i statement of LCO 3.6.1.2 is appropriate if the containment leakage rate limits are exceeded. The proposed action statement will continue to not permit entry into Mode 4 (i.e.,

reactor coolant temperature above 200 F) if the conditions of j LCO 3.6.1.2 are not met. The restrictions on containment leakage rates for entry into Mode 4 will be contained in Specification 6.17.

The proposed revisions to SR 4.6.1.2 will remove the prescriptive testing and scheduling requirements from this

' surveillance requirement. Instead, this surveillance requirement will require that containment leakage rates be determined in accordance with the Containment Leakage Rate Testing Program. The Containment Leakage Rate Testing Program will be based on the guidelines contained in RG 1.163 as required by Specification 6.17. RG 1.163 endorses NEI 94-01

' as an acceptable method for complying with the provisions of option B in Appendix J to 10 CFR 50. In addition, NEI 94-01

. references ANSI /ANS-56.8-1994 for detailed descriptions of the technical methods and techniques for performing Type A, B, and C tests. Since the Containment Leakage Rate Testing Program will be based on NRC endorsed guidance to implement the i provisions of option B in Appendix J to 10 CFR 50, the specific reference to Appendix J of 10 CFR 50 along with the l test frequencies can be deleted from SR 4.6.1.2.

The proposed deletion of the current SR 4.6.1.2.a, b, c, d, f and g, which contain specific details on test schedules, test

. accuracy verification, test methods and error analyses, is

consistent with the ISTS. The ISTS does not contain this level of detail concerning containment leakage rate testing.

The specific guidance on conducting containment leakage testing is contained in option B of Appendix J directly or in B-4

, ATTACHMENT B, _continu d Proposed Technical Specification Change Nos. 223 and 94 )

Page 5 l reference documents which are endorsed by the NRC. Therefore, it is not necessary to specifically state these requirements .

in.the technical specifications. The proposed SR 4.6.1.2.a {

will reference that Type A, B and c testing, except for l containment air lock testing, will be performed in accordance l with the Containment Leakage Rate Testing Program. The l current SR 4.6.1.2.e has been generally incorporated into the proposed SR 4.6.1.2.b. The wording has been modified by deleting the words " demonstrated OPERABLE." This wording is

{ not necessary since LCO 3.6.1.3 requires that each air lock be demonstrated operable. Since SR 4.6.1.2 specifically pertains to containment leakage rates, it is more appropriate to reference the testing portion of SR 4.6.1.3 only. j I

The proposed revision to general note number (3) of LCO !

3.6.1.3 titled " Containment Air Locks" will ensure that both l action statements are entered if air lock leakage results in '

exceeding the combined containment leakage rate. The proposed action statement of LCO 3.6.1.2 is applicable during Modes 1 thru 4 and specifically addresses containment leakage rates.

The proposed addition of the words " acceptance criteria" will i clarify the intent of this note. {

The proposed revisions to SR 4.6.1.3 will remove the I prescriptive scheduling requirements from this surveillance f requirement. Instead, this surveillance requirement will be performed at the frequency specified in the containment Leakage Rate Testing Program. Since NEI 94-01, one of the bases documents for the Containment Leakage Rate Testing Program, prescribes guidance on the required test frequency for containment air lock testing, the proposed change will continue to ensure that containment air lock testing is l performed at the test frequency endorsed by the NRC. This l change is consistent with ISTS, in that, the ISTS does not contain specific details on containment air lock testing frequency. The proposed elimination of footnotes (7) and (10) will allow the details contained in these two footnotes to be contained in Specification 6.17 titled " Containment Leakage Rate Testing Program." Therefore, this information will be applicable to all leakage rate testing performed in accordance with the Containment Leakage Rate Testing Program including containment air lock testing. The remaining two footnotes will be renumbered to reflect the deletion of footnote (7).

The proposed revision to the current footnote (9) will reflect that specific acceptance criteria will no longer be contained in LCO 3.6.1.2. The other changes to SR 4. 6.1. 3 will add consistency to this surveillance requirement and are necessary due to elimination of the specific test frequencies.

The proposed revision to the action statement of LCO 3.6.1.6 titled " Containment Structural Integrity" is necessary to reflect appropriate action should the structural integrity of the containment be found not to meet the LCO acceptance B-5

- _ _ . _ . - _ _ . ~ . . _ . _ . . _ _ . _ __

1 .

ATTACHMENT B', continutd

. Proposed Technical-Specification Change Nos. 223 and 94 i Page 6l I criteria during Modes 1 thru 4. The proposed wording will i continue to not permit entry in Mode 4 (i.e., reactor coolant temperature above 200*F) if the conditions of LCO 3.6.1.6 are not met. This change'is consistent with the proposed action

![

1 statement of LCO 3.6.1.1, LCO 3.6.1.2 and ISTS.

i The proposed revisions to SR 4.6.1.6.1 Will remove the i prescriptive scheduling requirements from this surveillance 4

requirement. Instead, this . surveillance requirement will be performed at the frequency specified in the containment J Leakage Rate Testing Program. Since RG 1.163, one of the l bases documents for the Containment Leakage Rate Testing

Program, prescribes guidance on the required test frequency i for containment vessel inspections, the proposed change will

continue to ensure that this inspection is performed at the frequency endorsed by the NRC. The proposed change to the i acceptance criteria is consistent with the acceptance criteria j for visual inspection of the containment stated in Option B of i 10.CFR 50 Appendix J. This change is consistent with ISTS, in

that, the ISTS does not contain specific details on

< containment vessel inspection frequencies.

{ The proposed revision of SR 4.6.1.6.2 will remove the

prescriptive reporting requirements from this surveillance l requirement. Instead, this surveillance requirement will

! require reports to be prepared in accordance with the Containment Leakage Rate Testing Program. Since NEI 94-01, l one of the bases documents for the Containment Leakage Rate Testing Program, prescribes the requirement for report l

preparation, the proposed change will continue to ensure that

reports will be prepared as required by the NRC. This change

! is consistent with ISTS , in that, the ISTS does not contain

! specific details on containment visual inspection reports.

! Administrative controls Section 6.9.2.g is no longer required

due to changes to NRC reporting requirements. The final rule i- on 10 CFR Parts 50, 55, and 73 titled " Reduction of Reporting

! Requirements Imposed on NRC Licensees" dated March 14, 1995 (60 FR 13615) no longer requires that containment leakage rate i tests summary reports be submitted to the NRC. Since the report of the visual inspection of the containment vessel is part of containment leakage rate tests summary reports, this

reporting requirement can be deleted.

The proposed addition of Specification 6.17 titled

" Containment Leakage Rate Testing Program" will ensure that a program is established which meets the requirements of 10 CFR l 50, Appendix J, Option B. This specification also states the

- values for Pa and L. as required by 10 CFR 50, Appendix J, Option B. The proposed wording for Specification 6.17 will 3

state the modified containment leakage rate acceptance 4

criteria previously contained in LCO 3.6.1.2 as discussed in the justification for changes to LCO 3.6.1.2. The current

j. B-6

ATTACHMENT B, continu;d Proposed Technical Specification Change Nos. 223 and 94 Page 7 exemptions to Appendix J of 10 CFR 50 are contained in proposed footnote (1). These exemptions are currently denoted in the containment systems LCOs. Since these LCOs will no longer contain specific details on containment leakage testing, it is necessary to move this information to proposed Specification 6.17. Proposed footnote (2) is required to allow a plant startup with a containment isolation valve (s) inoperable. Specification 6.17 requires that for the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria be based on a maximum pathway leakage rate (MXPLR). With one of two series isolation valves inoperable, for example, the MXPLR may be based on the inoperable valve. Therefore, the leakage rate acceptance criteria could be exceeded which would prohibit entry into Mode 4. Footnote (2) allows the MXPLR to be determined for these cases on the isolation device leakage rate and not on the inoperable valve leakage rate. The proposed listing of isolation devices on Footnote (2) is consistent with LCO 3.6.1.1. The proposed wording states that Specification 4.0.2 does not apply. The requirements of 10 CFR 50, Appendix J, Option B may be violated if the surveillance extension provided by Specification 4.0.2 were used. Guidance on frequency extension should be based on the requirements of 10 CFR 50, Appendix J, Option B and not under the generic requirements of Specification 4.0.2. This addition is consistent with ISTS. The proposed wording ,

clarifies that the provisions of Specification 4.0.3 applies.

The application of Specification 4.0.3 is consistent with the current technical specifications concerning containment leakage testing and the ISTS. The Index is required to be revised to reflect the addition of Specification 6.17.

l The changes to the Bases Sections 3/4.6.1.2 and 3/4.6.1.3 are necessary due to the proposed changes to their respective !

Specifications. Editorial changes are also included to provide consistency between the Specification wording and the Bases wording and to provide additional examples of appropriate Action statements to be entered depending on I equipment availability. l D. SAFETY ANALYSIS The proposed change will not affect the ability of the containment to provide a fission product barrier following a Design Basis Accident (DBA). The containment leakage rate i will continue to be determined using NRC endorsed test I methodologies and guidance on test frequency which have been determined to demonstrate that the containment will limit leakage to the value assumed in the accident analysis I following a DBA. The containment leakage rate assumed in the accident analysis ensures that offsite dose consequences does not exceed 10 CFR Part 100 limits following a DBA.

B-7 l

1 i

ATTACHMENT B, continued l Proposed Technical Specification Change Nos. 223 and 94 l Page 8 The proposed amendment does not change the required test pressure (Pa) for conducting Type A, B, and C testing. The maximum allowable "as found" overall Type A leakage rate will 1 be slightly increased for BVPS Unit No. 1 only, but will not j exceed the value assumed in the accident analysis. The "as left" measured overall Type A containment leakage rate and the measured combined Type B and C leakage rate limits will not be increased. Therefore, allowable containment leakage rate ;

limits, for Mode 4 entry, will not be increased. The maximum allowable "as found" overall Type A leakage rate will remain ;

unchanged for BVPS Unit No. 2.

The ability of the containment air locNs to provide a fission product barrier remains unchanged. The containment air locks will continue to be tested in a manner which will demonstrate their ability to perform this function. The proposed changes do not lower the required test pressure for conducting containment air lock testing. The maximum allowable containment air lock leakage limit remains unchanged.

The containment vessel will continue to be inspected at a frequency which will demonstrate that the structural integrity of the containment vessel is being maintained. Reports on the visual inspection will continue to be prepared in accordance with the Containment Leakage Rate Testing Program.

The proposed deletion of the reporting requirement 6.9.2.g is administrative in nature and does not affect plant safety.

The proposed addition of Specification 6.17 will ensure that the Containment Leakage Rate Testing Program is based on a periodic testing program which has been determined to be adequate to verify the leakage integrity of the containment and those containment systems and components which penetrate the containment. The remaining changes are editorial in nature and do not affect plant safety.

Therefore, this change is considered safe, based on the fact that the revised Specifications will continue to require adequate testing be conducted on a periodic basis to demonstrate the ability of the containment to provide a

. fission product barrier following a DBA. The "as left" measured overall Type A containment leakage rate limit will continue to provide margin between measured containment leakage and the containment leakage rate assumed in the accident analysis for calculating offsite dose consequences.

The proposed change will not impose additional challenges to the containment structure in terms of peak pressure. The calculated offsite dose consequences of a DBA will remain unchanged since the assumed containment leakage rate and the maximum allowable "as found" overall Type A containment leakage rate are equal.

B-8

l

, ATTACHMENT B, continugd Proposed Technical Specification Change Nos. 223 and 94 Page 9 E. NO SIGNIFICANT HAZARDS EVALUATION The no . significant hazard considerations involved with the proposed amendment have been evaluated, focusing on the three standards set forth in 10 CFR 50.92(c) as quoted below:

The Commission may make a final determination, pursuant to the procedures in paragraph 50.91, that a proposed amendment to an operating license for a facility licensed under paragraph 50.21(b) or paragraph 50.22 or for a testing facility involves no significant hazards consideration, if operation of the facility in accordance with the proposed amendment would not:

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

The following evaluation is provided for the no significant hazards consideration standards.

1. Does the change involve a significant increase in the probability or consequences of an accident previously j evaluated? !

I Containment leakage is not an accident initiator. The i proposed amendment does not add or modify any existing plant equipment. Therefore there is no increase in the !

probability of an accident previously evaluated.

. The consequences of an accident previously evaluated are l

not significantly increased. The proposed changes do not j affect the assumptions, parameters or result of any i Updated Final Safety Analysis (UFS A.") accident analyses. j The containment leakage rate will continue to be j maintained within the limit assumed in the accident !

analysis for a Design Basis Accident (DBA). The proposed j i changes do not modify the response of the containment l during a DBA. The proposed amendment will continue to j ensure that the ability of the containment structure, ;

including the containment air locks, to limit leakage i from a DBA is demonstrated using test methodologies and i

guidance on test frequencies that have been determined to be acceptable to meet the requirements of 10 CFR 50, Appendix J, Option B.

i The potential increase to overall accident risk due to the !

containment leak tightness decreasing between extended j l

B-9 !

3 t

t

)

ATTACHMENT'B,ccontinu:d- _

l Proposed Technical Specification Change Nos. m and 94 i Page 10 j i- testing intervals and the resulting potential increase.d j i radioactivity release to the environment during a DBA has l l

been determined to . be minimal based on th e findings of l

] NUREG 1493 titled " Performance-Based Containment Leak-Test Program." In addition, ' due to the performance based j nature of 10 CFR 50 Appendix J, Option B, the extended j 1 test intervals are . utilized only when the component (s) '

, have demonstrated an acceptable performanca history.

Therefore, a significant decrease in containment leak ;

^

tightness between extended test. intervals is not expected as a result of this proposed change.

l I Based on the above discussion, it is concluded that this

change does not involve a significant increase in the j' probability or consequences of an accident previously 4 evaluated.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? ;

h The proposed change does not involve any physical changes ,

i to the plant or changes in plant operating configuration. ]

l The proposed amendment involves changes to plant programs ,

i and administrative requirements used in determining i acceptable containment performance. The performance of plant systems, including the containment structure, during i plant operation remains unchanged.

Therefore, the proposed change does not create the

possibility of a new or different kind of accident from '

any accident previously evaluated.

i-

}- 3. Does the change involve a significant reduction in a j margin of safety?

il The margin of safety is not significantly reduced by this proposed change. The acceptance criteria for "as left" measured containment leakage rates is not being increased as result of this proposed amendment. For Beaver Valley l Power Station (BVPS) Unit No. 1 only, the "as found" maximum allowable overall Type A leakage rate is being slightly increased. However, the slight increase does not

, exceed the value assumed in accident analysis for

' containment leakage during a DBA due to changing the acceptance criteria from less than to less than or equal i

to. The margin between the acceptable "as left" measured i overall Type A containment leakage' rate and the leakage <

L rate. assumed in the accident analysis is not being I

decreased.

The maximum "as found" allowable overall Type A leakage i rate remains unchanged for BVPS Unit No. 2. The margin B-10 I

\

b c ATTACHMENT B, continuad Proposed Technical Specification Change Noc. 223 and 94 Page 11 between the acceptable "as left" measured overall Type A containment leakage rate and the leakage rate assumed in the accident analysis is also not being decreased.

The maximum allowable measured combined Type B and C leakage rate is not being increased above the ' current limits.

The maximum ' peak ' containment pressure following a DBA remains unchanged. The containment depressurization time following a DBA remains unchanged. The calculated'offsite I dose consequences of a DBA remains unchanged. l The proposed amendment continues to ensure reactor )

containment- system reliability by periodic testing in compliance with 10 CFR 50, Appendix J, Option B. The extension of Type A, B and C test frequencies permitted by 10 CFR 50 Appendix J, Option B, is not expected to result in a significant decrease in containment leak tightness between test intervals. Due to the performance based nature of 10 CFR 50 Appendix J, Option B, the extended '

test intervals are utilized only when the component (s) have demonstrated an acceptable performance history.

Therefore, a significant decrease ~ in containment leak tightness between extended test intervals is not expected as a result of this proposed change. l The changes which are either administrative or editorial in nature will not reduce the margin of safety because they have no impact on any safety analysis assumptions.

. Therefore, based on the above discussion, it can be concluded that the proposed change does not involve a

- significant reduction in a margin of safety.

j F. NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION l 1

, Based on the considerations expressed above, it is concluded

! that the activities associated with this license amendment I

request satisfies the no significant hazards consideration i standards of 10 CFR 50.92(c) and, accordingly, a no significant hazards consideration finding is justified.

l

+

B-11

ATTACHMENT 7-1 Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Change No. 223 w;.

Applicable UFSAR Changes a

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BVPS-1-UPDATED FSAR Rav. 13 (1/95)

REGULATORY GUIDE 1.144, SEPTEMBER 1980: AUDITING OF QUALITY ASSURANCE PROGRAMS FOR NUCLEAR POWER PLANTS Beaver Valley Power Station - Unit 1(BVPS-1) will meet the intent of Regulatory Guide 1.144 for the auditing of its Quality Assurance Program during the operations phase with the following clarifications and alternatives:

Paracraoh C.1 The applicability of the referenced regulatory guides / ANSI standards (RG 1.28: ANSI N45.2, RG 1.28:

ANSI N45.2.9, and RG 1.74: ANSI N45.2.10) is as stated in the respective positions on these regulatory guides / ANSI standards as described in the UFSAR.

Paracraoh C.3 Scheduled internal audit frequency will be as specified in Paragraph C.? of Regulatory Guide 1.33, Rev. 2, February 1978.

The pre-audit and post-audit conferences required by Sections 4.3.1 and 4.3.3 of ANSI N45.2.12-1977 may be fulfilled by a variety of communications such as telephone conversations.

REGULATORY GUIDE 1.155, JUNE 1988: STATION BLACKOUT The utilization of BVPS emergency diesel generators as alternate AC (AAC) power sources for coping with station blackout, and the reliability program for these g Regulatory Guide 1.155 (June 1988' .yggrgyprs follow the_ guidance of O mm"d 1.3.4.2 American National Standards Institute (ANSI) Standards N45.2.5: DRAFT 3, REVISION 1, JANUARY 1974, " SUPPLEMENTARY QUALITY ASSURANCE REQUIREMENTS FOR INSTALLATION, INSPECTION AND TESTING OF STRUCTURAL CONCRETE AND STRUCTURAL STEEL DURING THE CONSTRUCTION PHASE OF NUCLEAR POWER PLANTS" The Duquesne Light Company follows the guidance of ANSI N45.2.5, Draft 3, Revision 1, January 1974. Procedures and/or specifications were developed prior to, and implemented concurrent with the start.of the operations phase.

N45.2.8: DRAFT 3, REVISION 2, SEPTEMBER 1973, " SUPPLEMENTARY QUALITY ASSURANCE REQUIREMENTS FOR INSTALLATION, INSPECTION, AND TESTING OF MECHANICAL EQUIPMENT AND SYSTEMS FOR THE CONSTRUCTION PHASE OF NUCLEAR POWER PLANTS" The Duquesne Light Company follows the guidance of ANSI N45.2.8, Draft 3, Revision 2, September 1973. Procedures and/or specifications were developed prior to, and implemented concurrent with the start of the operations phase.

1.3-51 (Pr.p.s.) A rc

0 0 fb - \ - u P O ATED T'5 A R TN SE R T "1" RG No. \.)63 ; Sep 4embt 171r uvs A A RJenna s.chen- 5.G Pub m..scens.3 (mA.. A L.4-TM 6,.m u

h t. bnkoswktsb ltcNogt . Re.k Thsksg [e.go~ O th orcodentt w A k A gusJelsnes conb..wed iw RG No. 1.16 3 . N rp dduy buid' pmo,Jo g u , A . , te. o , o, occepgi ,_

p,J o e s.,t _ 6.iel le.R bd p rege. , leuf r.k 4ok m,Ak.4 3 , pre cebecs ; ad amolp es AV my b s. w e d A o c pl y w,kh N pedermem b.3e3 og hm 6 in Apf. J,y 7 oE lo C Fit 5' O .

1 l

Ytoposeh W erbs

_ . . _ . _ _ _ _ . . . _ . _ .. ______.___-__.m _. . .

BVPS-1-UPDATED FSAR Rev. 10 (1/92) l

3. .The nuclear control operator utilizing remote indicating control room instruments determines the containment atmospheric ambient pressure, temperature and radiation j levels.

i

4. These levels are compared to predetermined levels which '

specify the protective apparatus and allowable times for

, the entering personnel.

5. Personnel entering the containment- are provided with necessary instruments to verify the radiation levels,

particulate activity levels and oxygen content of the j containment atmosphere prior to proceeding to the work '

j

. area.

1 6. Any significant changes of containment atmospheric

!- parameters noted by an room or any plant condition annunciator in the main control i

which, in the opinion of the nuclear control operator or shift supervision, could pose

! -a threat to the safety of personnel in the containment

! will require the evacuation of personnel from the

containment.

~ Tests and Insoections 1

The- containment vacuum ejector is not considered a part of the '~

engineered safety features and, since it is such a simple mechanical device, periodic tests are. not required.

i i

mechanical containment vacuum pumps are operated during The the initial containment leakage rate test (Section 5.5) and demonstrated to have adequate capacity to remove inleakage.

3 During normal unit operation, they are alternated in service, thus

)

providing periodic testing of each containment vacuum pump.

5.4.2.2. Containment Leakage Monitoring System 4

Desian Bases

, The containment leakage monitoring system is used to determine the

leakage rate of the containment under periodic test conditions.

The containment leakage rate is determined using the absolute test

! method, and either the !?as s Point or Total Time data analysis 1

method is used to calculate the containment leakage rate.

1 The h

(bpenc. less system provides for measurement of containment leakage rate of than 0.1 percent of the contained volume in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months with an 3 uracy sufficient to meet the requirements of Appendix J, 10 CFR 2

) 50 The system is designed in accordance with ANSI N45.4, j

i American National Standard, Leakage Rate Testing of Containment Structures for Nuclear Reactors, March 17, leakage rate testing 'is 1972. Containment

' Appendix J with conducted in accordance with 10 CFR 50, 1

certain exceptions as noted in the Technical Specifications. '

(bp hen h 5.4-10

BVPS-1 UPDATED FSAR Rsv. 8 (1/82)

The ' environmental conditions (atmospheric pressure, temperature, and humidity) inside and outside the containment structure were continuously monitored during the test to evaluate their contribution 4

j-

to the response of the containment. The test was not conducted under i

extreme weather conditions such as snow, heavy rain, or strong wind.

1 containment structure was subjected was to the peak test ,

! When the expected to be t internal pressure, the maximum radial growth maximum vertical approximately three-quarters inch and the

(

deformation at the dome apex to be approximately one and one-half l inch. These deformations were calculated for the analytical stress

' evaluation of the containment liner. Strain measurements were made strain gages at adequately on the steel liner using conventional 1

i selected points.

During the acceptance test, visual examination and instrumentation in measurements, both were used to record cracking and changes vertically and radially, due to the response of the concrete Prior

containment structure to the air pressure test of the liner.

structure was surveyed, to testing, the outside of the concruce l measured, and inspected for cracks, ard all pertinent information During the test, measurements were made of the radial j

recorded.

deflections at various locations on the smil from the top of the mat i

to the spring line of the dome. Twa permanent pits located approximately 90 degrees apart were provided for access to the l

I containment wall below ground grade. Thase pits allowed localized visual inspection and measurements of the lower part of the wall.

Vertical deflections were measured at the apex and spring line of the i

dome. Additional strain measurements were made on the surface areas adjacent to the equipment access hatch and in other areas where i

l l

stresses were critical.

were measured by linear variable differential Deformations (LVDT's) mounted at the internal surface of the linear

)

transducers l

LVDT's were also used to measure displacements of the

! plate. Cracks larger than 0.01 concrete ring around the equipment hatch.

which occur during the test were recorded. They were measured l inch After the by an optical comparator and checked with feeler gages.

completion of the test recovery of the structure was recorded. The i

l crack pattern was again inspected and recorded.

The containment concrete surface was whitewashed in areas of high

! chart crack patterns. Photographs were stress and at openings to taken of the crack patterns to provide permanent records.

I Temperature, barometric pressure and weather conditions were recorded l hourly during the test period.

Containment Leakaae Rate Tests

! The containment leakage rate tests are performed in accordance with j the guidelines of Appendix J f 10CFR50, " Primary Reactor Containment

! Leakage Testing for Water Cooled Powar Reactors".

I hbon 5.6-3 l

Yrojabe

l

i BVPS-1 UPDATED FSAR Rsv. 8 (1/90) i f

i The containment leakage testing program includes the performance of Type A tests, to measure the containment ovu all integrated leakage

]

i rate, Type B tests, to measure leakage of certain containment 3

components, and Type c tests, to measure containment isolation valve

]

leakage rate.

1 The preoperational Type A test was conducted according to the rules of Section III. of A pendix J g in accordance with 6 ;tien III.O. Q

Periodic tests / are conductedexemption the Technical

! Appendix J J (with the noted performed in using the leakage i specifications). These tests are 4 monitoring system (described in Section 5.4.2.2).

The measured leakage rate does not exceed the design basis accident t

l hours of the weight of leakage rate (La) of 0.1 percent per 24 l containment air at the calculated peak containment pressure of 40.0 psig. The remaining leakage characteristics are determined in

{

(64. 6hme um,y keh T=ti:n: !!!.A.4.0 2nd III.A.S. '. @ e 6 accordance withta;;:ndix L S j % t'ryr.~ U.L ATFS j Type B tests are carried out to a6nitor tne principal sources of leak l development in accordance with Appendix J, Sectica III.B.1 Teet Methods. These tests are performed to measure leakage originating at l

containment penetrations, air lock door seals, equipment and i

j personnel access hatches, and all other components which may develop leaks and require repairs to meet the acceptance criterion of the Type A test. Qwy preoperational and-peciecMejType B tests conducted according I The to the rules of Appendix J f C::tien !!!.0.2 by local pneumatic i

pressurization of the containment components at a pressure not less than Pa. The acceptance criterion fora Type B_ tests is given in 1;; ndi J, S: tie = TII.E. g tu c LATfy p .4 g The periodic Type B tests are scheduled according to the guidelines I

l of App;ndix J, Cectimi III.D RT Q The Type C ' tests are performed on the isolation valves to verify their sealing capability and leaktightness according to Appendix J, S :tien III.C.1. The test includes valve closure and leakage tests.

A valve closure test is conducted prior to a valve leakage test to r demonstrate the proper sealing capability of a valve upon which are receipt normally of an isolation signal. Those isolation valves and sealing capabilities.

1 closed are exercised to verify closure Those containment isolation valves which are in a system that is expected to be filled with water for 30 days following a LOCA and therefore do not represent a containment atmosphere leak pathAppendix are not

,~

subject to the Type C test requirements of 10CFR50 J . !!' . M l

2 cA 60 4

i !

5.6-4 l

ktopoJc) Wer

BVPS-1 UPDATED FSAR R;v. 8 (1/82)

The Type C tests are conducted according to the guidelines ofL-Appendix J, Sectien III.O.14 by local pneumatic pressurization at a pressure Pa. gg T~"

periedim Type v cesta are conducted ;;;srding te ti.. geldelines F Appendix 7, S:;tien III.D.3. DELETE T The structural integrity of the containment will be determined during the shutdown for each Type A containment leakage rate test in 4 accordance with the Technical Sp:;ificet ens.43cgTr{}

5.3-la, " Containment Isolation Arrangements", lists the Table containment isolation valves which can be individually leak tested.

9 4

4 j

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i 1 l l

1 i

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5.6-5

((mf ased War p l

__. -- . _ . . _ . - -.-.. _..-_. - . .--.-. .- . _- __ _ _ .- __ ._- . . . . . . - - _ ~

i i ATTACHMENT C-2

! Beaver Valley Power Station, Unit No. 2 Proposed Technical Specification Change No. 94 <

4 i

! Applicable UFSAR Changes t

t I

l 4

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b d

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)

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I i

i 1

I I.

i e

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t I l t

l I,

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I I.

. BVPS-2 UFSAR Rev. 6 TABLE 1.8-1 (Cont) j RG No. 1.150, Rev. 1 UFSAR Reference Section 5.3.1, 5.3.3 ULTRASONIC TESTING OF REACTOR VESSEL WELDS DURING PRESERVICE AND INSERVICE EXAMINATIONS iFEBRUARY 1983)

Ultrasonic testing of the reactor vessel welds during preservice and

! inservice examinations at BVPS-2 will follow the guidance of this regulatory guide as described in the Preservice Inspection Program, which was submitted to the NRC in Letter 2NRC-5-154, dated December

26, 1985, and the Inservice Inspection Program, which is scheduled to be submitted to the NRC in the last quarter of 1986.

RG No. 1.155. June 1988 UFSAR Reference Section 8.3.1.1.19 STATION BLACKOUT BVPS utilites the emergency diesel generators at each unit as an alternate AC (AAC) power source to operate systems necessary for coping with a station blackout. The design of the cross-tie circuit between BVPS-1 and BVPS-2 AAC power sources conforms with guidance provided by RG No. 1.155.

A00-V i

(s No. \,\G S.,\es.r MQ txF5 fad ke..Ircrenco. Seehoe b'3'b

d t w 6.sel Co A , w A L eu;7 34 , tem u ,, ,

le f.m ' s iw oeceed... it i

DC (ordomwed Le.Q.g Rede. To l

u,, W h g u,At\ w e co h tJ m (( G IJo , b \ 6 3 .

l a CA on k s ble pf tbotAenet- f cattey h3 h \ e.ko r gusbe (J/*M'3 yu d*u he se.) \..k hed frgr~, Idy f *h* N' "O*0' i fk *c'O *"') l CI' W' b ptr o/ N ^cA So3khd sq h e. uld Ib Co^p}

5 O Ae, t

ogi rm 6 m g j,x7 of 16CFit50, - ,

I l

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79 of 80 Y(%fo3 l l

1

d BVPS-2 UFSAR 4

computer system. Indicators are provided in the main control room to monitor hydrogen gas concentrations and an indicating light shows the hydrogen recombiner is operating. A recorder for hydrogen gas concentration (channel A only) is provided.

The following controls and instruments are located on the hydrogen analyzer panel: a stream selector switch for stream to be analyzed,

, indicating lights for reference /zero gas pressure, or calibration / sample gas pressure low alarm, and high gas concentratioa.

!' The hydrogen recombiner inline heater is controlled from its own control panel in the safeguards area. When in AUTO (heater on i permissive), the heater will energize if the hydrogen recombiner blower is running and if the metal temperature of the heater is less than 625'F.

6.2.6 Containment Leakage Testing

The containment leakage rate tests are performed in accordance with 10 CFR 50, Appendix J,1%M. ag GDC 52, 53, and 54.

GSplsew1U The purpose of the containment leakage test program is to assure that

., leakage through the reactor containment, systems, and components penetrating the containment boundary does not exceed the allowable

leakage rate values as specified in the Technical Specifications

]. (Chapter 16) or other design base documents. -

f The containment leak testing program includes the performance of Type A tests to measure the containment overall integrated leak rate; Type B tests to detect local resilient seal leakage at electrical penetrations, equipment hatch, personnel hatch, emergency escape

trunk, and fuel transfer tube flange; and Type C tests to measure containment isolation valve leakage rates.

6.2.6.1 Containment Integrated Leak Rate Test - Type A

0fLETE The 6- eper-tirnr1 m>B periodic Type A leakage rate test will be 4

conducted in accordance with 10 CFR 50, Appendix J, Per : sph !!.#0]fh$sma8-) .

Protest requirements will be identified and included as part of the

~ Type A test procedure to ensure that the necessary preparations,

precautions, and temporary modifications have been completed prior to i Type A test commencement. Such pretest requirements will include unit status, instrumentation requirements, support systems status, i temporary test or measurement equipment requirements, supplementary i

a l 6.2-72 (fre g ed We d

i

BVPS-2 UFSAR testing requirements, general containment inspection requirements prior to containment closeout, personnel assignment, shift briefings, TW tet vbeut. w h~' b Cod mnd L4=%y Mc Tg),) 9c p ((,,LR 79)]

P r ir . ;; c .._ ;. c. ...: ef the Type ^ ::: , a general inspectio'n of the accessible interior and exterior surfaces of the containment structure will be performed for the purpose of identifying evidence of deterioration which may effect the containment structural integrity or leaktightness. Visual inspection will be performed to detect and observe: gross deformations of the interior surfaces of steel containment liner; paint failure due to massive rusting, electrolysis, or abrasion; evidence of exterior concrete spalling or cracking; high stress areas of the containment concrete such as

?

f I

i 4

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4 6.2-72a i

Rsv. 2 BVPS-2 UFSAR equipment hatch, personnel hatch, electrical and valve penetration areas; accessible areas at the bend line; shake space integrity, etc.

Should evidence of containment degradation be found, the Type A or structural acceptance test will not be performed until an evaluation I has been performed and repairs made, if required. Such structural 1 deterioration and subsequent corrective actions taken will be reported g pert of _t h e Tyn A tcat re.;;;rt described in Sectier.

9 % occorhome W M h c LRW,)

System Venting and Draining To place the primary reactor containment system as close to post-accident conditions as possible, those portions of the fluid systems that are part of the reactor containment boundary that may be opened diractly to the containment or outside atmosphere under post-accident conditions will be opened or vented to the appropriate atmosphere during the test.

Those lines which are normally fluid-filled and which may be drained or have the fluid driven off by the accident, including portions of l systems inside or outside containment that penetrate the containment and may rupture as a result of a LOCA, will be drained to the extent necessary to expose the containment isolation valve seats to the l containment atmosphere, except as noted by the following. Systems I that are required for proper conduct of the test or to maintain Cd BVPS-2 in a safe condition during the test shall be operable in their normal mode and need not be vented or drained. Additionally, systems pro d O M

M that are normally filled with water and operable under post-accident _ eyM*4 Of conditions, such as the CHRS, need not be vented or drained.6 Systems nd u of f that are not vented or drained during the Type A test and which could g d become exposed to the containment atmosphere during a leakage DBA !

will be Type C tested, and the Type C test leakage rate for the pode- 'oM l

penetration path will be added to the upper confidence limit, i The tect pressure to which the containment is subjected during the l Type A test is equivalent to the calculated peak containment pressure following the design basis accident. Temporary air compressors will be utilized to raise containment pressure. When the containment has j reached test pressure, containment temperature will be monitored for '

a period of not less than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months until stabilization criteria have been met. Once stabilized, the containment parameters of temperatuce, pressure, and vapor pressure will be observed and recorded ter the duretion of testing. The duration of the test l period will be sufficient to enable adequate data to be accumulated and analyzed so that s leakage rate and upper confidence unit can.be I accurately determined. During this period, the containment leak rate l will be calculated by the mass point or total time analysis technique I to verify that it is within the limits of the BVPS-2 Technical Specifications re<;uirements . Upon uetermination of an acceptable leakqe rate, a W rification test.wil. Le performed to confirm the l capability of t'ae method and the test instrumentation used to i determine the c.catainment !

l l

6.2-73 3

t' % n\*C

. . . - - - - - . . = . . - - -. - _ _ - . .- .

e i BVPS-2 UFSAR leakage rate. Having met all test criteria, the containment will be vent and raduced to atmospheric conditions.

cd le(@

The4 acceptance criteria for an acceptable leakage rate test requires that containment leakage be less than 0.75 La, as defined by 10 0""

l *a, J.rp;;dir J. A superimposed leak test will be conducted

< immediately following the Type A test. The results from this test will be considered acceptable nrovided the differ:::: 5:tu::: th:

superimpo leak test data and-** w"g: 1 t :t det: is ler tP--

0.25 ' ^. w AW Abe acetows. Ag4 :TntU m ANf1] ANI 59 IWC ke perledle teet fail: te Acht the act:pt=:: ::it::i: p::ified, the USF".C rill be ::tified rd th: Typ: A t::t ::hed 1 ;pplicchie to cubrequent tect: vill 5: di:: :::d eith the US""C. If tuo

D'Bf7M c^--ecutive peri ~81c Type i lee'a:e t--te feil te reet the applicebis accept ree eriterie, e Type A lee'c; teet eill be p: fere:d at rech

fuelin; :hutd- - :: every le :::the, rhichever :::ur: first, =til tee ernretutive Type i lerks;e te-te ::t the acerpt er: criteria, after which : :::::1 ::t::t ::h:d=1: elli th:2 h: :::=:d.

i 6.2.6.2 Containment Penetration Leakage Rate Test - Type B 4

l Type B containment penetration leakage tests are conducted in Type B leakage dhe d.M,J accordance withM9 C"" 59, i;;:rdir J, S::ti: !!!.".

/ tests are intended to detect local leakage and to measure leakage across containment electrical penetrations, equipment and personnel hatches, emergency escape trunk, and fuel transfer tube flange.- A list identifying all containment penetrations is provided in Table 6.2-60.

The makeup air method of testing, which will primarily be used to measure Type B leakage, consists of the pressurization of a component with air or nitrogen and measuring leakage using a flowmeter installed in the pressurization line.

The test pressure to which Type B tests will be conducted is identical to that specified in Section 6.2.6.1 for Type A testine.

In: pavindi- retect- e9edule f:: Typ: Bt::::gN;__y. ;_: :: rr) ,

dee- net ranntra nerfnr=ane== A" ring ::ch r-=-+ar c hu t d~-r fer i

fu;1ing, but in na e ee t et intervele ;-eater th e 2 y;:::. Air l 1eche chall he tested at 6 rnth intervele. M eever, air leck- rhen epened dur ; such intervel: eill 5: t::ted :ft:: :::h :p::ing ::::pt i

ich:2 th: cirl::h i: being :::d f:: : ltiple ;;tri::, ch:2 it :h:14-bee i L*e=*-A e - per ? deye.

l i

The periodic retest schedule for Type B testing will be in accordance I with Li^ """ L M io ,

7......yh III.O.2.

D LL AW.3 Containment Isolation Valve Leak Rate Tests - Type C i

6.2.6.3 1 l

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6.2-74 Ytsposb Uok m

i 0- BVPS-2~UFSAR

, +.

Type C testing'is1 performed on containment isolation valves to verify their sealing capability and leaktightness. All testing will be performed in accordance with the requirements ofg10 C.R 50, f.;; din h CLNIfh i

I l

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1i 6.2-74a 1 1

i fo geS c dom 4

s l BVPS-2 UFSAR OELE'I p

v h _ " - - - - -" " ' rGv4) ' Typ e C tests will be performed by local

pressurization applied in the same direction as that when the valve would be required to perform its safety function, unless it can be demonstrated that testing in a reverse direction is as conservative.

Each valve to-be tested will be closed by its normal means, that is, motor, solenoid, diaphram, handwheel, etc, and will receive no additional adjustments (hand-tightening after closure by motor) or preliminary exercising.

The containment isolation valves will be tested by local pressurization to the pressure specified in Section 6.2.6.1 for the Type A test. The test method will be to vent and drain a system, or portions thereof, and to pressurize across one, or a series of valves with air or nitrogen using primarily the makeup air method described in Section 6.2.6.2. Test connections located on both the inlet and outlet sides of a valve, or pair of valves, are provided to facilitate system draining and/or pressurization. Leakage will be measured using an installed flow meter in the pressure supply line.

y er On multiple valve penetrations, only the highest leaking valve shall 45 leh be recordea as uneV penetration leak rate. Valves, and their respective system status which must be Type C tested, are listed in Table ,6.2-60. Test vents, drains, and connections located between isolation valves will have two barriers (valve with cap, and valve with flange) and will be administrative 1y controlled. These connections will not be leak tested.

The test pressure will be as specified in Section 6.2.6.1 for Type.A testing.

The acceptance criteria for allowable leakage associated with Type B and Type C combined leakages is to be in accordance 'eith in OER 10, Appendi.x J, Peregraph: !!!.B.3 and !!!.CA . g Scheduling for eachperiodicTypeCtestwillbeinaccordancewithd 10 Cr" 50, Appendix J, " crc,e eph III.O.2, cad pcrformed dur4ng each reester chutde:- for refueling, but in nc :::: ct inter /cle grectu then ? ya m .

6.2.6.4 Scheduling and Reporting of Periodic Tests eM1.ETE _

c Ot!LET E-The C-epce3 schedules for h ^^erational a @. periodic tests are in l accordance with W-CIE '^ ^ppendix 1 < _ l

.W CLAT h. l 1- 0^a** went integrated lech rcte (Type A): " ccperat-ienal ;

Type ! teeting vill be perferred pe4er te fuc1 Iced. !

Periedic Type A tecting vill be performed three ti-es at equal in*arvels during a 10 year pe r4cd. The third tect will be perforwed when ""PC 2 is shut de n fer the 10-year plant in-cer" ice inspectia= ,

2- C0000in : t penetratica leakago LosL1 % (T g g,

,soopoIO&I a nc! Type B t0 ting vill bc Ocmp10ted pp4 x 4

~

lor Fo Ac M A g 6 d C hi wy

. [ pc.per.how 3

_a.wm2-n (ProgxAWorch e __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - - _ _ _ - __ . _ _ _ _ _ _ - _ . ___--___ _ _____

g BVPS-2 UFSAR s

f;;l 1 :d :nd vill 5: n plet: t: suppert the Type A cest.

Periedi: Typ: 2 t:: ting, ether then cir teche, vill be perf::::d durin; 22:5 ::::ter :: fueling chutd:er er : her

,nnu.n4.ne incoru.t., un, 4n om og.e et i- t e rg,.1 g . 3, e ,.

th-- 2 yeer . Air lecke vill be perledi: lly t;:::d :very 5-c:nthe :: cft:r Openings.

3. Contain;;nt isolation velv. looke a. Lonina (Tyy. C);

Dreeperatienel Type C testing vill he ::;picted pri : te fuel lead --d 'cill he cerplete te suppert the Type A testf Periedic Type c teeting cill be perfereed duria: each

ft: ling chutdren, but in n: :::: et intervel gre:ter than 2 y re. F.eperting of preepere+1^nel and pa-4aA4r Tyra= A, n, .-a c t.gt4 u111 h. <n .ormra nr. ut+h in cru sn, Appendi- 1, Parag-apb= U R 1 9, and 9 6.2.7 Fracture Prevention of Containment Pressure Boundary Materials A summary of the fracture toughness characteristics of the containment pressure boundary materials and the confirmation of compliance to GDC Sri can be found in the DLC transmittal to the NRC (Woolever 1983).

6.2.8 References for Section 6.2 Aerojet Nuclear Company (ANC) 1976. RELAP4/ MOD 5: A Computer Program for Transient Thermal Hydraulic Analysis of Nuclear Reactors and Related Systems. User's Manual Vol I-III, Report ANCR-NUREG-1335.

American National Standard: L_ t itut: laSI. Containment System Leakage Testing Requirements. ANSI /ANS-56.8- (This document used only as a guideline.) 9 American Nuclear Society (ANS) 1978. Decay Heat Power in Light Water Reactors. ANS Proposed Standard 5.1, Revised September 1978.

Anderson, T M. (Westinghouse) 1979. Personal Communication (Letter NS-TMA-2075 dated April 25, 1979) to J.F. Stolz, USNRC. Westinghouse LOCA Mass and Energy Release Model for Containment Design -

March 1979 Version.

Bloom, G. R., et al. 1982. Hydrogen Distribution in a Containment with a High Velocity Hydrogen-Steam Source, presented at the Second International Workshop on the Impact of Hydrogen on Water Reactor Safety, Albuquerque, New Mexico.

Bordelon, F.M., Massie, H.W., Sr., Zordan, T.A. 1974a. Westinghouse Emergency Core Cooling Evaluation Model Summary. WCAP-8339.

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l 6.2-76 !

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