Text

Proposed Tech Specs 1.1 Re definitions,3.6.1.1 Re Primary Containment - operating,3.6.1.2 Re Primary Containment Air locks,3.6.1.3 Re PCIVs,3.6.5.1 Re Drywell & 5.5 Re Programs & Manuals
	ML20134D825
Person / Time
Site:	Perry
Issue date:	01/31/1997
From:	; CENTERIOR ENERGY
To:	;
Shared Package
ML20134D817	List: ML20134D825; PY-CEI-NRR-2133, Application for Amend to License NPF-58,requesting Implementation of App J,Option B,Performance-based Primary Containment Leakage Rate Testing Program for Type A,B & C Leak Rate Tests;
References
	NUDOCS 9702050395
	Download: ML20134D825 (24)
	v • d • e

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Attachment 3 Definitions

. PY-CEl/NRR-2133L

!

Page lof 8 1,1 i

1.1 Definitions (continued)

EMERGENCY' CORE COOLING SYSTEM (ECCS) RESPONSE TIME The from when ECCS RESPONSE TIME shall be that tim the monitored initiation setpoint at tie>arameter channel exceeds sensor its ECCS until the ECCS equipment is ca>able of performing its

safety function (i.e.. tie valves travel to their required positions, pump discharge pressures reach their required values etc.).

Times shall include i

i diesel generator starting and sequence loading delays where applicable. The response time may

be meas,ured b overlapping, y means of any series of sequential.

' or total steps so that the entire response time is measured. Exceptions are stated ll in the individual surveillance requirements.

END OF CYCLE RECIRCULATION PUMP TRIP The E0C-RPT SYSTEM RESPONSE TIME shall be t (E0C-RPT) SYSTEM RESPONSE time interval from initial movement of the associated turbine stop valve or the turbine TIME

! ' control valve to complete suppression of the i electric arc between the fully open contacts of i

the recirculation ) ump circuit breaker. The 1 i

response time may .)e measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. .

4 ISOLATION SYSTEM

' RESPONSE TIME The ISOLATION SYSTEM RESPONSE TIME shall b i

. time interval from when the monitored parameter exceeds its isolation initiation setpoint at the channel sensor until the isolation valves travel l to their required positions. The response time may be measured by means of any series of s, !

sequential,over the entire respo.

~

lapping,'or total steps so that nse time is measured. Exceptions 3 are stated in the individual surveillance requirements.

f (L,

The rate.maximum L allowable rimary containment leakage air weig,. shall be 0.2 % of primary containment ht per day at the calculated peak containment pressure (P.). f (continued)

PERRY - UNIT 1 1.0-3 Amendment No. 59, 77 g 2 SDOS Nobs 440 p PDR w .

Attachment 3 PY-CEl/NRR-2133 L Primary Containment-Operating Page 2 ors

, 3.6.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.1.1 Perform required visual examinations and 6---NOTE------

leakage rate testing except for primary SR 3.0.2 is not

, containment air lock testing, in applicable i accordance with 10 CFR 50, A;;;r. dix J, e-----------

~

r /

h,v s n n77e. n. u. n. A. a..vamn s .e m..n. A. 4. f 4. a. A.

_ , -t. i. n. n. .e .

i /The leakage rate acceptance criterion is with 10 CFR 50, I

$ 1.0 L,. However, during the first unit ^ppendiv J, as l startup following testing performed in mediried by accordance with 10 CFR 50, Appendix J, as appmund modified by approved exemptions, the exa ptiens-leakage rate acceptance criteria are J l

< 0.6 L for the Type B and Type C tests,) l Land < 0,.75 L. for the Type A tcst. y J

r C

-the ?nnwr3 odoanment Leo.% Pcde ~I6shog Pagrun, N- l l

1 l

./

PERRY - UNIT 1 3.6-2 Amendment No. 69

Attachment 3 PY-CEl/NRR-2133L Primary Containment Air Locks Page 3 of 8 3.6.1.2 I SURVEILLANCE REQUIREMENTS l

SURVEILLANCE FREQUENCY SR 3.6.1.2.1 ------------------NOTES------------------

1. An inoperable air lock door does not i invalidate the previous successful

. performance of the overall air lock leakage test. fle, Ltc 4,

2. During MODES , 2, and 3, results shall be eva uated against acceptance criteria S SR 3.6.1.1.1 n accordance with 10 CFR 50, d { Appendix J, as modified by approve (exemptions. _f I

Perform required primary containment air --NOTE------

lock leakage rate testing in accordance with 10Trn ou, Appendix J, as modifiedl 63.0.2isnot SR applicable b-----------M y approved exemptions.

,,_, The acceptance criteria for air lock In acc_ordance

.<' testing are: l 0 CFR 50,

(.. } -

ppendix J, as

a. Overall air lock leakage rate is modified by 5 2.5 scfh when tested at 2 P,. approved (xemptions

b. For each door, leakage rate is s 2.5 scfh when the gap between the or seals is pressurized to 2 P,h i

SR 3.6.1.2.2 Verify primary containment air lock seal 7 days air header pressure is 2 90 psig.

i the Smaq ConhanWN L&tA9c W (continued) hbnj Proyetn,

)

PERRY - UNIT 1 3.6-7 Amendment No. 69

Attachment 3

- PY-CEl/NRR 2133L Page 4 ors PCIVs 3.6.1.3

^ )- SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.9 ------------------NOTE-------------------

Only re 3

and 3. quired to be met in MODES 1, 2.

) Verify the combined leakage rate for all secondary containment bypass leakage

---NOTE------

paths is s 0.0504 L, when pressurized to ISR3.0.2isnot a p,, applicable q..............)

In accordance g withfl0 CFR 50.

dix J. as

(-%_PnnwgCmbnnud~ W Me ,pp g by

. Tes-bg3 fhpm , xemotion

.)

SR 3.6.1.3.10 ------------------NOTE-------------------

Only re and 3. quired to be met in H00ES 1, 2.

Verify leakage rate through each main ---NOTE-----

steam line is s 25 scfh when tested at

= P, . Until the end of Operating Cycle 6, [SR 3.0.2 is not the leakage rate through one main steam applicable line at = is P limited to s 35 scfh when tested as long as the total leakage rate t------------J l through,all four main steam lines is-s 100 scfh. In accordance

. with 1.0 CFR 50W Appendix J, l

as modified by approve C ennwq (,n%nmed-teWr O Mg i gxemption1)d W avrn _ A (continued)

~

PERRY - UNIT 1- 3.6-18 Amendment No. 83

I i *

i ,

Attachment 3 PY-CEI/NRR-2133L PCIVs i

' Page 5 of 8 3.6.1.3 i

_ SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE i FREQUEllCY 1

J l SR 3.6.1.3.11 1 ------------------NOTE-------------------

4 only required to be met in MODES 1, 2, 1 -

and 3.

( Verify combined leakage rate of I gpm V 3

times the total number of PCIVs through SR----NOTE----h 3.0.2 is

! hydrostatically tested lines that 1

penetrate the primary containment is not , ' not applicable a i t----------- j-- i exceeded when these isolation valves are -

tested at 1 1.1 P,.

' In accordance dx

j. nag %md skg b W ha*pp d Pq/3rm , exemptions I

i SR 3.6.1.3.12 ---------------- NO TE------- -----------

4 1

~_. ...) Only required to be met in MODES 1 -

I 2, and 3. j i

Verify ~each outboard'42 inch primary 18 months

containment purge valve is blocked to ,

restrict the valve from opening > 50'.

i i

SR 3.6.1.3.13 N 1

NOT E-------------------

Mot required to be met when the Backup -

Hydrogen Purge System isolation valves -

are open for pressure control, ALARA or air quality considerations for personnel

- entry, or Surveillances or special l testing of the Backup Hydrogen Purge System that require the valves to be i open. 1

_ ______________________________________ i t

! Verify each 2 inch Backup Hydrogen Purge 31 days System isolation valve is closed. ;

i i

1 PERRY - UNIT 1 3.6-19 Amendment No. 85

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

s Attachment 3 i PY-CEl/NRR-2133L Drywall Page 6 of 8 3.6.5.1 i

j SURVEILLANCE REQUIREMENTS j SURVEILLANCE FREQUENCY SR 3.6.5.1.1 Verify bypass leakage is less than or ------NOTE-----

equal to the bypass leakage limit. The performance j However, during the first unit startup of the trywell following bypass leakage testing bypass leakage performed in accordance with this SR, the test is i acceptance criterion is s 10% of the extended to the drywell bypass leakage limit. sixth refueling outage and need not be i

performed !

j during the i j fifth refueling !

l outage.

1 18 months SR 3.6.5.1.2 Visually inspect the exposed accessible kceprior

interior and exterior surfaces of the performa of i

drywell. each T eA ,

tes equired ;

b ,

j 3.6.1.1.1. l

/ l i

bree times during each i

10-year service period, at approximately equal rvais.

j l 4

i PERRY - UNIT 1 3.6-44 Amendment No. M/,82

R-2133L Programs and Manuals Page 7 or 8 5.5 5.5 Programs and Manuals l

- I 5.5.10 Safety Function Determination Proaram (SFDP) (continued) 1 The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, j

} the appropriate Conditions and Required Actions of the LC0 in '

which the loss of safety function exists are required to be entered.

5.5.11 Technical Specifications (TS) Bases Control Proaram 3

) This program provides a means for processing changes to the Bases for these TS.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC cpproval provided the changes do not involve either of the following:

1. a change in the TS incorporated in the license; or

2. a change to the USAF: o Bases that involves an unre. viewed safety cues'. ion as defined in 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with tce USAR. l

d. Proposed changes that meet the criteria of Specification 5.5.11.b.1 or Specification 5.5.11.b.2 above shall be reviewed and approved by the NRC prior to f& implementation. Changes to the Bases implemented without y

prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

.h k

E Inser ew Section 5.5.12 Primary Containment Leakage Rate Testine P ingram here (see following page) l PERRY - UNIT 1 5.0-15 Amendment No. 69

. NEW SUBSECTION 5.5.12 TO BE INSERTED AT THE END OF PAGE 5.0-15 FOR TECHNICAL SPECIFICATION SECTION 5.5 Attachment 3 RR-2133L PROGRAMS AND MANUALS [,Y 5.5.12 Primarv Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the primary containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B as 1 modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, " Performance-Based Containment Leak-Test Program," as modified by the following exceptions:

I BN-TOP-1 methodology may be used for Type A tee.

l The corrections to NEl 94-01 which are identified on the Errata Sheet attached to '

the NEl letter, " Appendix J Workshop Questions and Answers," dated March 19, 1996, are considered to be an integral part of NEl 94-01.

The peak calculated primary containment internal pressure for the design basis loss of coolant accident is P . l l

The maximum allowable primary containment leakage rate, L , shall be 0.20% of l primary containment air weight per day at the calculated peak containment pressure (P ).

Leakage rate acceptance criteria cie;

a. Primary containment leakage rate acceptance criterion is < 1.0 L,. However, during the first unit startup following testing performed in accordance with this Program, the leakage rate acceptance criteria are < 0.6 L, for the Type B and Type C tests, and s 0.75 L, fur the Type A tests;

b. Air lock testing acceptance criteria are:

1) Overall air lock leakage rate is s 2.5 scfh when tested at 2 P,.

2) For each door, leakago rate is s 2.5 scfh when the gap between the door seals is pressurized to 2 P3.

The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Primary Containment Leakage Rate Testing Program.

The provisions of SR 3.0.3 are applicable to the Primary Containment !.eakage Rate Testing Program.

[The remainder of the pags in Section 5.0 should be renumbered, as necessary.]

. - . . - . - - . - - - ~ ~ --- - - - * ^ ~ - ~ - - -^

-~~~~~~^~~~"~}

Attachment 4 I. .

PY-CEI/NRR-2133L TABLE 0F CONTENTS rage i or i6 f

l B 3.9 REFUELING OPERATIONS (continued)

B 3.9.2 Refuel Position One-Rod-Out Interlock B 3.9-5 B 3.9.3 Control Rod Position . . . . . . . . . . }

s B 3.9.4 Control Rod Position Indication .......... . . . . . . . B 3.9-9 i B 3.9.5 Control Rod OPERABILITY-Refueling . . . . . . . . . B 3.9-13 5 B 3.9.6 . B 3.9-16 l

Reactor Pressure Vessel (RPV) Water Level-i i

B 3.9.7 Irradiated Fuel . . . . . . . . . . . . . . . . . . B 3.9-19 i i

Reactor Pressure Vessel (RPV) Water Level- !

New Fuel or Control Rods ............. B 3.9-22 !

4 B 3.9.8 Residual Heat Removal (RHR)-High Water Level j B 3.9.9 ..... B 3.9-25 !

Residual Heat Removal (RHR)-Low Water Level . . . . . .

B 3.9-30 I

3.10 . SPECIAL OPERAT

~

3.10.1 Inservice Luk and Hydrostatic Testing Operation . . . . 3.10-1 4

3.10.2 Reactor Mode Switch Interlock Testing .........

3.10.3 Single Control Rod Withdrawal-Hot Shutdown ......

3.10i4 s 3.10.4 3.10 6 !

) Single Control Rod Withdrawal-Cold Shutdown . . . . . . 3.10-9 j 3.10.5 Single Control Rod Drive (CRD) Removal-i 3.10.6 Refueling . . . . , . . . . . . . . . . . . . . . . 3.10-13 Multiple Control Rod Withdrawal-Refuelin 3.10-16 j 3.10.7 Control Rod Testing-0

3.10.8 SHUTDOWN MARGIN Test-Refueling (SDM) perating

........ . . 3.10-18

...g.......

......... 3.10-19 B 3.10 SPECIAL OPERATIONS !

\

B-3.10.1 Inservice Leak and Hydrostatic Testing Operation . . . . B 3.10-1

, B 3.10.2 Reactor Mode Switch Interlock Testing ......... B 3.10-6 B 3.10.3 Single Control Rod Withdrawal-Hot Shutdown ...... B 3.10-11 i B 3.10.4 Single Control Rod Withdrawal-Cold Shutdown . . . . . . B 3.10-16 i

! B 3.10.5 Single Control Rod Drive (CRD) Removal-t Refueling ..................... B 3.10-21 B 3.10.6 Multiple Control Rod Withdrawal-Refueling . . . . . . . B 3.10-26 B 3.10.7 Control Rod Testing-O B 3.10.8 SHUTDOWN MARGINTest-Refueling (SDM) perating......... ............. B 3.10-29 B 3.10-33 4,0 DESIGN FEATURES i 4.1 Site Location ............ 4.0-1 4.2 Reactor Core . . . . . . . . . . . . . . . . . . . . . 4.0-1 4.3 Fuel Storage . . . . . . . . . . . . . . . ......

4.0-2 5.0 ADMINISTRATIVE CONTROLS i 5.1 Responsibility . . . . . . . . . . . . . 5.0-1 i 5.2 Organization . . . . . . . . . . . . . . . . . . . . . 5.0-2 5.3 Unit Staff Qualifications . .... 5.0-4 5.4 Procedures . . . . . . . . . . . . ........... . . . . . . . . . . 5.0-5 5.5 Programs and Manuals . . . . . . . ......... 5.0-5.6 Reporting Requirements . . . . . . . . . . . . .,. . . 5.0 t 'l 5.7 High Radiation Area .. . .............. 5.0 20 PERRY - UNIT 1 vii Revision No. 0

.- = - -. _

SR Applicability Attachment 4 B 3.0 PY-CE!/NRR-2133L Page 2 of 16 SR 3.0.2 The 25% extension does not significantly degrade the (continued) reliability that results from performing the Surveillance at its specified Frequency. This is based on the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the Thererore, when a test SRs. The exceptions to SR 3.0.2 are those Surveillances for 1:terval is specified in the Which the 25% extension of the interval specified in the regulations, the test inters al Frequency does not apply. These t ndividual Soecifications/peations An example are stated of where SR in 3.0.2 cannot be extended by the ~

TS, and the Surveillance oes not apply is a Surveillance With a Frequency of "in Requirements will then aCCordance With 10 CFR 50. Accendix J._ as modified by include a NOTE in the roved exemntions "/ The reouirements of regulations take frequency stating asn 3.0.2 edence ove' r the TS.g/f.he TS achfifiot in ana or themselve is not applicable." A Awnu a L6E miervai specified in the regulations.

example or an exception Therefore, there is a Note in the Frequency stating.

when the test intervalis not "SR 3.0,2 is not applicable."

spectried in the regulations is the statement in the As stated in SR 3.0.2, the 25% extension also does not apply Primary Containment to the initial portion of a periodic Com)letion Time that Leakage Rate Testing requires performance on a "once per..." ) asis. The 25%

Program that "the extension applies to each performance after the initial provisions orsn 3.0.2 do performance. The initial performance of the Required not apply..." This exception Action, whether it is a particular Surveillance or some ~

is provided because the other remedial action. is Considered a single action with a

, Program already includes single Completion Time. One reason for not allowing the 25%

extension to this Completion Time is that such an action (extension or test intervals.

A usually verifies that no loss of function has occurred by checking the status of redundant ar diverse components or accomplishes the functior, of the inoperable equipment in an alternative manner.

The provisions of SR 3.0.2 are not intended to be used repeatedly merely as an operational convenience to extend Surveillance intervals (other than those consistent with refueling intervals) or periodic Completion Time intervals beyond those specified. ;

SR 3.0.3 SR 3.0.3 establishes the flexibility to defer declaring affected equipment inoperable or an affected variable outside the spccified limits when a Surveillance has not been completed within the specified Frequency. A delay aeriod of up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or u tha^quency, Tre whichever is less,p applies from thetopoint the inlimit time of the sp l

it is discovered that the Surveillance has not been periormed in accordance with SR 3.0.2. and not at the time (continued)

PERRY - UNIT 1 B 3.0-12 Revision No. O

NcSN.u33L Primary Containment-Operating Page 3 of 16 B 3.6.1.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.1 Primary Containment-Operating -

r c #y;I (" ^

t o s'de h" "

BASES

^1 (ine s) -

BACKGROUND The function of th primary containment is to isolate and System following al,esign Basis Accident (DBADa oy confine the postulated release of radioactive material to within limits. The primary containment consists of a free standing steel cylinder with an ellipsoidal dome, secured to

' a steel lined reinforced concrete mat, which surrounds the Reactor Coolant System and provides an essentially leak tight barrier against an uncontrolled release of radioactive material to the environment. Additionally, this structure provides shielding from the fission products that may be present in the primary containment atmosphere following-accident conditions.

The isolation devices for the penetrations in the primary containment boundary are a part of the primary containment leak tight barrier. To maintain this leak tight barrier:

a.

All primary containment penetrations required to be closed during accident conditions are either:

1.

capable of being closed by an OPERABLE containment automatic isolation system, or primary

2. closed by manual valves, blind flanges, or de-activated automatic valves secured in their closed positions, except as provided in LCO 3.6.1.3. " Primary Containment Isolation Valves (PCIVs)";

b.

Primary containment air locks are OPERABLE. except as 3rovided in LC0 3.6.1.2, " Primary Containment Air

_ocks":

c. The equipment hatch is closed and sealed:

d. The leakage control systems associated with 3enetrations are OPERABLE. except as provided in

.C0 3.6.1.8, "Feedwater Leakage Control System." and LCO 3.6.1.9, " Main Steam Isolation Valve (MSIV)

Leakage Control System (LCS)";

(continued)

PERRY - UNIT 1 B 3.6-1 Revision No. 1

Attachment 4 PY-CEI/NRR.2133L Primary Containment-Operating Page 4 of 16 B 3.6.1.1 BASES c

4 BACKGROUND e.

(continued) The containment leakage rates are in compliance with the requirements of Specification 3.6.1.1 and Specification 3.6.1.3:

f. The suppression pool is OPERABLE: and g.

The sealing mechanism associated with each primary containment penetration, e.g. , welds, bellows, or 0-rings, is functional.

This Specification ensures that the performance of the primary containment, in the event of a DBA, meets the assumptions used in the safety analyses of References 1 and 2. SR 3.6.1.1.1 leakage rate requirements are in conformance with 10 CFR 50. Appendix J Ref. 3) as modified by approved exemptions, onP2)

APPLICABLE SAFETY ANALYSES The safety design basis for the primary containment is that it must withstand the pressures and temperatures of the limiting DBA without exceeding the design. leakage rate.

The DBA that postulates the maximum release of radioactive material within primary containment is a LOCA. In the analysis of this accident, it is assumed that primary i containment is OPERABLE such that release of fission products to the environment is controlled by the rate of primary containment leakage.

Analytical methods and assumptions involving the primary I containment are presented in References 1 and 2. The safety analyses assume a nonmechanistic fission product release following a DBA. which fonns the basis for determination of offsite doses. The fission product release is, in turn, based on an assumed leakage rate from the primary containment. OPERABILITY of the primary containment ensures that the leakage rate assumed in the safety analyses is not exceeded.

The maximum allowable leakage rate for the primary containment (L ) is 0.20% by weight of the containment and drywell air pe,r 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at the ximum peak containment--

pressure (P ) of 7.80 psig (Ref. .

des n %se LocA)

Primary containment satisfies Criterion 3 of the NRC Policy 2

Statement.

PERRY - UNIT 1 B 3.6-2 Revision No. 1

NnNa.u33L Primary Containment-Operating z, Page 5 of 16 B 3.6.1.1 BASES (continued)

{

v LC0 Primary conuainment OPERABILITY is maintained by limiting leakage to6D1.0 L., except prior to the first unit startup bnPrimary accordance Containment f te er performing a required 10 CFR 50, Appendix J leakage with the p(sty At this time, theAombin0d Typ m

Leakage Rate Testing <y_ye v.~ u

< 0.0 Q ,M ty merg,., ,.ype n, ,ea i u t,bc Programj containm,. Compliance with this LCO will s en @ure a primary ent configuration, including the equipment hatch, app h bk that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analysis.

rnd Individual leakage rates specified for the primary 44 containment air locks are addressed in LC0 3.6.1.2.

APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES.

Therefore, primary containment leakage limits are not required to be met in MODES 4 and 5 to prevent leakage of radioactive material from primary containment.

(refer tc o.6.1.10. " Primary Containment-Shutdown").

ACTIONS .A l In the event that primary containment is inoperable, primary containment must be restored to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Comaletion Time provides a period of time to correct the pro)lem that is commensurate with the importa'nce of maintaining during MODES 1, 2, and 3. primary containment OPERABILITY This time period also ensures that the probability of an accident (requiring primary containment OPERABILITY) occurririg during periods when primary containment is inoperable is minimal.

B.1 and B.2 If wit 3rimary containment cannot be restored to OPERABLE status 11n the associated Com)letion Time, the plant must be brought to a MODE in whic1 the LCO does not apply. To achieve this status, the plant must be brought to at least MODE allowed Com3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The experience,pletion Times are reasonable, based on operating to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

J (continued)

?ERRY - UNIT 1 B 3.6-3 Revision No. 1

$N-2133L Primary Containment-0perating ;

Page 6 of 16 B 3.6.1.1 BASES (continued) l SURVEILLANCE SR 3.6.1.1.1 REQUIREMENTS Maintaining the primary containment OPERABLE requires compliance with the visual examinations and leakaan rate i i

te fritm g #(modifiec test recuirements ofII{f LtK 60. ADDendix J (Ref. 3). aW by approved exemptionsJ Failure to meet air lock !

gnd tenhqe,

^

leakage testing (SR 3.6.1.2.1 and SR 3.6.1.2.4). secondary

& Tg pa<p. S containment bypass leakage (SR 3.6.1.3.9), resilient seal primary containment purge valve leakage testing (SR 3.6.1.3.6), main steam isolation valve leakage (SR 3.6.1.3.10). or hydrostatically tested valve leakage (SR 3.6.1.3.11) does not necessarily result in a failure of this SR. The impact of the failure to meet these SRs must be evaluated against the Type A. B. and C acceptance b oprovea i exemptions f The treauency is required bYcriteria D CFR 50. Appendix J. as modified by approved exemptions.

(Thus. SR 3.0.2 (which allows Frequency extensions) do W

I The Appendix J fopw Q '

Additionally. Lexemptions approved to date are listed below.

Bechtel Topical Report BN-TOP-1 may be i

utilized for ILRTs with a duration of less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in accordance with Section 7.6 of ANSI N45.4-1972 (Reference 5).y

a. Section III.D.2(b)(ii) - The air lock seal leakage test of Section III.D.2(b)(iii) of Ap)endix J may be substituted (following normal air loct door opening)

[ for the full-pressure test provided that no maintenance has been performed that would affect the air locks sealina capability (Reference 6).

I I

{[ Note: This is performed per SR 3.6.1.2.

w

b. Section III.D.3 - A one time schedular Exemption was issued to permit Type C testing of certain containment isolation valves to exceed the two year interval, so that these tests could by conducted during the first refueling outage (Reference 7).

(continued)

[for Appendix J, Option B testing, unless they have bee

((Reference 3). _

PERRY - UNIT 1 B 3.6-4 Revision No. 1

. Attachment 4

ry-crimRR.2133L Primary Containment-Operating Page 7 of 16 B 3.6.1.1 BASES i 1

SURVEILLANCE I i SR 3.6.1.1 (continued) I REQUIREMENTS c.

Sections III.A.1(d) III.A.5(b)(2) III.B.3 and III.C.3 - The main steam lines between the inboard and outboard MSIVs (including the volume up to the i 1

outboard MSIV before seat drain line valves) are not

, recuired to be vented and drained for Type A testing anc ;

the main steam line isolation valve leak rates are !

exempted from inclusion in the overall integrated primary containment leak rate and the combined local leak rate (Reference 8). !

d. i Section III.D.1(a) - The third Type A test for each I 10-year service period is not required to be conducted l when the plant is shutdown for the 10-year plant 4 inservice inspection (Reference 8).

e.

4 Section III.D.3 - Type C local leak rate testing may be performed at other convenient intervals in addition to shutdown during refueling, but at intervals no greater than 2 years (Reference 8).

j As left leakage 3rior to the firs i I

a re tartup after performing be < quired (10 CH 60, Appendijx J leakage test is required to 0.75 L for overall Type A leaka e0.6 L for combined Type

@J lbetween , required leakage es rate e t all other times is based on an overall Type A eakag ance criteria e, , A W1.0 L, the offsite dose consequences are-bot de 70 T. . At S 1 by the !

assumptions of the safety analysis.

g l

REFERENCES 1. USAR Section 6.2.

2. USAR. Section 15.6.5.

3. 10 CFR 50 Appendix Plenf B

4. PY-CEI/NRR-1510L, dated June 24, 1992.

5. Letter from NRC (B.J. Youngblood) to CEI (M.R.

Edelman), " Performance of the Preoperational Containment Integrated Leak Rate Test - Perry Nuclear Power Plant Unit 1." dated June 10, 1985.

(continued)

PERRY - UNIT 1 B 3.6-5 Revision No. 1

attacamen,4 Primary Containment-Operating PY-CEl/NRR-2133L Page 8 or 16 B 3.6.1.1 BASES REFERENCES (continued) 6.

PNPP Safety Evaluation Report Supplement 7. Section

6.2.6 " Containment Leakage Testing." November 1985.

7.

Letter from NRC (T. Colburn) to CEI (A. Kaplan). '

" Exemption from 10 CFR Part 50. Appendix J". dated January 22. 1988.

8. Letter from NRC (J. Ho Company (D. Shelton). pkins) to Centerior Services

" Issuance of Exemption from the Requirements of 10 CFR Part 50. Appendix J - Perry

~

Nuclear Power Plant. Unit 1". dated December 4, 1995.

9. Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program.

10. NEl 94-01, Revision 0, " Industry Guideline for implementing Performance-Based Option of 10 CFR Part 50, Appendix J, including the Errata Sheet attached to the NEl letter, l

" Appendix J Workshop Questions and Answers," dated March 19,1996.

s i

i PERRY - UNIT 1 B 3.6-6 Revision No. 1

. Attachment 4

. PY-CEl/NRR 2133L primary Containment Air Locks Page 9 of 16 B 3.6.1.2 BASES ACTIONS D.1 and D.2 (continued)

If the inoperable 3rimary containment air lock cannot be restored to OPERAB_E status within the associated Completion Time while operating in MODE 1, 2. or 3. the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE allowed Com 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The experience,pletion Times are reasonable, based on operating to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

E.1 and E.2 If the inoperable )rimary containment air lock cannot be 4

restored to OPERAB_E status within the associated Completion Time during operations with a potential for draining the reactor vessel (0PDRVs). or during movement of recently irradiated fuel assemblies in the primary containment.

action is required to immediately suspend activities that l represent a potential for releasing significant amounts of radioactive material, thus placing the unit in a Condition that minimizes risk. If applicable, movement of recently irradiated fuel assemblies in the primary containment must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also. if applicable, action must be immediately initiated to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Action must continue until DPDRVs are suspended.

SURVEILLANCE SR 3.6.1.2.1 REQUIREMENTS Maintaining primary containment air locks OPERABLE requires tt t Pri m t Ccabed comoliance with the leakage rate test requirements o L% P Amendix J (Ret. 2), as mod _ 1ed b a r N@* Mu-*'(10 CFR 50. Men in MODES

\axemotions This1.SR 2.reand 3.

lects the W:

leakage rate testing requirements with regard to air lock leakage (Type B leakage tests). The acce)tance criteria were established arior to initial air loct and primary containment OPERA 3ILITY testing. The periodic testing requirements verify that the air lock leakage does not exceed the allowed fraction of the (overal ary (continued)

PERRY - UNIT 1 B 3.6-14

p. [-hypc ~6 a JQ Revision No. 1

Attachment 4 PY-CEI/NRR-2133L Primary Containment Air Locks Page 10 of 16 B 3.6.1.2 BASES SURVEILLANCE SR 3.6.1.2.1 (continued) leakage rate. The Frequency is required by flcontainment,pphiun o, as moaniea oy approvea h ov. n Thus, SR 3.0.2 (which allows Frequency extensions) does not kPPIYdh PrimejT.mtnin%x uawy. we mq gep The Appendix J exemption related to air lock testing approved to date for PNPP is:

Section III.D.2(b)(ii) - The air lock seal leakage test of Section III.D.2(b)(iii) of Ap)endix J may be substituted (following normal air loc: door opening) for the full-pressure test provided that no main' tenance has been performed that would affect the air lock's sealing capability (Reference 5)

The SR has been modified by two Notes. Note 1 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 capable of providing a fission product barrier in the event of a DBA. Note 2 has been added to this SR. requiring results to be evaluated against the acceptance criteria SR 3.6.1.1.1 duri peration in MODES 1, 2. and 3.' Th s ensures tha air akage is properly accounted for in determini he vera rimary containment leakage rate.

b* a Since the( rima aff applicable in MOD '1, 2. containment leakage rate is only to 'and 3. the Note.2 requirement is imposedonlyduringlthese MODES.

SR 3.6.1.2.2 The Service and Instrument Air System pressure in the header to the primary containment air lock is verified to be at a 90 psig every 7 days to ensure that the seal system rcmains viable. It must be checked because it could bleed down during or following access through the air lock, which occurs regularly. The 7 day Frequency has been shown to be acceptable through operating experience and is considered adequate in view of the other indications available to operations personnel that the seal pressure is low.

(continued)

PERRY - UNIT 1 B 3.6-15 Revision No. 1

, .e..

Attachment 4

PY-CEl/NRR 2133L Primary Containment Air Locks Page 11 of 16 B 3.6.1.2 BASES SURVEILLANCE SR 3.6.1.2.3 REQUIREMENTS (continued)

The air lock interlock mechanism is designed to prevent simultaneous opening of both doors in the air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident primary containment support )rimarypressure (Ref. 3), closure of either door will ',

containment OPERABILITY. Thus, the interloc( feature supports primary containment OPERABILITY 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 will and that simultaneous not inadvertently occur. inner and outer door oaening '

Due to the nature of tnis interlock, challen and given that the interlock mechanism is only opened,ged when the primary containment air lock door is i this test is only required to be performed u>on entering or exiting a primary containment air lock aut is .

not required more frequently than once per 184 days. The 184 day Frequency is based on engineering judgment and is considered adequate in view of other administrative controls such as indications operations personnel. of air lock door status available to SR 3.6.1.2.4 A seal pneumatic system test to ensure that pressure does not decay at a rate equivalent to > 1.5 psig for a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from an initial leakage rate test to veri ressure system of 90 psig is an effective performance The 18 month Frequency is bas on the fact that operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency. which is based on the refueling cycle. Therefore. the Frequency was concluded to be acceptable from a reliability standpoint.

REFERENCES 1. USAR Section 3.8.

2. 10 CFR 50. Appendix pimh

3. USAR. Table 6.2-1.

4. USAR Section 15.7.6.

5. PNPP Safety Evaluation Report Supplement 7. Section 6.2.6 " Containment Leakage Testing." November 1985.

PERRY - UNIT 1 B 3.6-16 Revision No. 1 1

I Attachment 4 pCIVs PY-CEl/NRR-2133L Page 12 of 16 B 3.6.1*3 BASES SURVEILLANCE 3.6.1.32 (continued)

REQUIREMENTS exceed the times assumed in the DBA analyses. The Frequency of this SR is in accordance with the Inservice Testing Program. Additionally, the MSIVs must meet an average stroke time. This average stroke time shall be calculated using the stroke times of the fastest valve in each main steam line, and this average shall be a 3 seconds.

SR 3.6.1.3.8 Automatic PCIVs close on a primary containment isolation signal to prevent leakage of radioactive material from primary containment following a DBA or other accidents.

This SR ensures that each automatic PCIV will actuate to its isolation Josition on a primary containment isolation signal. T1e LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.1.5 overlaps this SR to provide complete testing of the safety function. The 18 month Frequency is Nsed on the need to perform this Surveillance under the cerations that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass this Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. l SR 3.6.1.3.9 l This SR ensures that the leakage rate of secondary 1 containment bypass leakage paths is less than the specified leakage rate. This provides assurance that the assumotion I in the radiological evaluations of Reference (f)are metT The leakage rate of each by] ass leakage path is aYsumed to be the maximum pathway leacage (leakage through the worse of the two isolation valves) unless the penetration is isolated by use of a closed manual valve, a closed and de-activated automatic valve, or a blind flange. In this case, the leakage rate of the isolated bypass leakage path is assumed to be the actual pathway leakage through the isolation device. If both isolaticn devices in the penetration are closed, the actual leakage rate is the lesser leakage rate of the two devices. This method of quantifying maximum h p;th-:y leakage rate is erly to be used for tM: SR (i.c.. L

.Sppc.idi" J mar 4 = patbey leakage rSte limit: are to bc t-quantificd in acccrdance ..ith tppendir J). G (continued)

PERRY - UNIT 1

'Qb' B 3.6-30 Revision No. 1 jog

,c..-

Attachment 4 py.CEI/NRR 2133L PCIVs page i3 or 16 8 3.6.1.3

BASES SURVEILLANCE REQUIREMENTS SR 3.6.1.3.2 (continued)

A Note is added to this SR which states that these valves are 1, 2 only and 3.required to meet this leakage rate limit in MODES In the other conditions, the Reactor Coolant Systemare liIpLits is not not pressurized and specific primary leaka e rate recuired. The Frecuency is required '

/CFR 50. AppencTX e (Ref. 4) as mocitied by appro exemptions: thus SR 3.0.2 (which allows Frequency xtensions) does not apply 7

h Pnmaq6nb%wk Ledge 12de Teshe Sgrw -

SR 3.6.1.3.10 jya g '

The analyses in References (2 and 3) re based on ' leakage that is less than the specified leakage rate.

each main steam line must be s 25 scfh when tested at PLeakage -

Until the end of Operating Cycle 6, the leakage through..one main steam line is limited to s 35 scfh when tested at = P ,

as long as the total leakage _. rate throuah all fout main-

. steam lines is 5100 scfh fThe MSIV leakage rate must be D (vsrifledtobeinaccordancewiththeleakagetest

) requirements of Reference 4. as modified by anrnvad

'yptemptions/The Frequency is recuired byffu L-4 50,.

t

pendix J (Rei. 4), as mod 171ec uy approved exemptions

tius, apply.SR 3.0.2 (which allows Frequency extensions) does'not

~

A only Note re is added to this SR which states that these valve are.

and 3. quired to meet this leakage. rate limit in MODES 1. 2, In other conditions, the Reactor Coolant System is not pressurized and specific

. rate limits are not required. primary containment leakage ihe frimq Gabnmed kdEe t O M9 *'

SR 3.6.1.3.11 Surveillance of hydrostatically tested lines provides assurance that the calculation assumptions of References 2 and 3 are met. The combined leakage rates must be' demonstrated at the frequency of-the-leakage test _

re uirements of/ Reference 4. as modifi~ea oy approvedW xemp lons: thus. SR 3.0.2 (which allows frequenc V xtensions) does not-apply r fiv Snartj Ccnknnwnt Irakaqc Rdc. T6hq Empra.(continued PERRY - UNIT 1 B 3.6-31 Revision No. 1 3

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

4 h.*

I Attachment 4 PY-CEI/NRR 2133L pCJy3 Page 14 of 16 B 3.6.1.3

) BASES SURVEILLANCE REQUIREMENTS SR 3.6.1.3.13 (continued)

Purge System (e.g., testing of the containment and drywell be open. ventilation radiation monitors) that require the valves to drywell purge valve requirements.The 31 day Frequency is REFERENCES 1. USAR. Chapter 15.

2. USAR. Section 6.2.

3. USAR. Table 6.2-32,

4. 10 CFR 50. Appendix hon 6) i I

1 l

.)

PERRY - UNIT 1 B 3.6-32a Revision No. 1

,e c.*

Attachraent 4 PY-CEI/NRR-2133L g '* y Page 15 of 16 BASES i

SURVEILLANCE SR 3.6.5.1.1 (continued)

) REQUIREMENTS i q least every nine months until two consecutive tests meet the limit at which time the 18 months Frequency may be resumed.

ypass This leakageSurveillance is less than or equal 2

ensures that the to the acceptable AN actual drywel value of 1.68 ft assumed in the safety analysis. As left drywell bypass leakage, prior to the first startup after performing a required drywell bypass leakage test, is !

required to be s 10% of the drywell bypass leakage limit.

At all other tines between required drywell leakage rate ,

tests, the acc_eptance criteria is based on design A/vT. At '

the design AME the containment temperature and '

pressurization response are bounded by the assumptions of ;

the safety analysis. The leakage test is performed every l 18 months, consistent with the difficulty of performing the l test, risk of high radiation exposure, and the remote aossibility that a component failure that is not identified ay some other drywell or primary containment SR might occur.

Operating experience has shown that these components usually aass the Surveillance when aerformed at the 18 month requency. Therefore, the requency was concluded to be acceptable from a reliability standpoint.

A Note has been 3rovided to modify the Frequency of this Surveillance. T1e Note reflects NRC approval of a one-time deferral of this test. from the fifth refueling outage to the sixth refueling outage.

SR 3.6.5.1.2 The exposed accessible drywell interior and exterior surfaces are inspected to ensure there are no apparent physical defects that would prevent the drywell from performing its intended function. This SR ensures that el3I^d) ,

drywell structural integrity is maintained. The Frequency y '

was)

}

drywelchosen so that the interior and exterior surfaces of the

((d,D%of the@ primary containment required by 10 CFR 50.be inspe Ng ,

Appendix Jhilcf. 2 e to the passive nature of the peuta u*ogm.'A"D (continued) hen the primary containment ins etions were placed onto a performance-based frequench

, the drywellinspections were retained at a frequency of 3 times in a 10-year inservice inspection period. The retention of this frequency was a commitment made to facilitate th placement of the Drywell Bypass Leak Rate Test onto a performance-based freque PERRY - UNIT 1 B 3.6-126 Revision No. 1

l !

,4 Attachment 4 - DryWell PY-CEl/NRR 2133L Page 16 of 16 B 3.6.5.1 BASES I 1

SURVEILLANCE REQUIREMENTS SR 3.6.5.1.2 (continued) drywell structure, the specified Frequency is sufficient to identify component structural integrity.degradation that may affect drywell l

1 REFERENCES 1. USAR, Chapter 6 and Chapter 15.

4. 10CFR50.,9p^ndixJ. #

I i

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I

~

I t .

l 1

1 l

I PERRY - UNIT 1 B 3.6-127 Revision No. 1

ML20134D825

Text

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