ML20108C943

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Proposed Tech Specs,Implementing 10CFR50,App J - Option B
ML20108C943
Person / Time
Site: Clinton Constellation icon.png
Issue date: 05/01/1996
From:
ILLINOIS POWER CO.
To:
Shared Package
ML20108C911 List:
References
NUDOCS 9605070216
Download: ML20108C943 (27)
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Attachment 3 to U-602565 LS-95-014 Page1of10 Marked-Up Pages from Facility Operating License No. NPF-62 and the Technical Specifications i

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i 9605070216 960501  :

PDR ADOCK 05000461 .I P PDR \

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(8) Post-Fuel Loadino Initial Test Procram fSectio'n 14. SER. SSER 5 and SSER 6)

Any changes to the initial test program described in Section 14 of the FSAR made in accordance with the provisions of 10 CFR 50.59 shall be reported in accordance with 50.59(b) within one month of such change.

(9) Emeroency Resoonse Canabilities (Generic Letter 82-33. Sucolement I to NUREG-0737. Section 7.5.3.1. SSER S and SSER 8. and Section 18.

SER. SSER S and Safety Evaluation Dated April 17. 1987)

a. IP in accordance with the commitment contained in a letter dated December 11, 1986, shall install and have operational separate power sources for each af the fuel zone level channels as provided for in Reguhtory Guide 1.97 prior to startup following the first refuel.ing outage.
b. IP shall, submit a detailed control room design final supple-mental summary report within 90 days of issuance of the full power license that completes all the remaining items identified in Section 18.3 of the Safety Evaluation dated April 17, 1987.

D. The facility requires exemptions from certain requirements of 10 CFR Part 50 and 10 CFR Part 70. These include: (a) an exemption from the requirements of 10 CFR 70_.24_for_ the criticalit alans n t rs around the f el tor e area '" " %I-+ '- - + - - " - - - - - +

, paraartph Ill.0.2(b)s.!f$f5^'g[5..d'i5'5,' Ab55 TIE 5 tie [U.5 5551le:k:;;

L te:t :t P: ef 2:ragraph III.D.2(b)(iii) fer th: :: tire cirleck te:t at -

(, P: Of p:ragr:pi III.O.2(b)(44-)-of .";;;;d4x J d:: :: =intentec- h J S::: perferr:d in the :f rleck th:t ce"ld :ffect it: :::lir.; ::pdility L (-Seetter 5.2.5 ef SSER 6); (c)$)an exemption from the re utrement of locF R Part 50,

- dixJ,Jexemptingtheadasu eakage~ rates Appm A J-cNare reph_ III.C.3 Of " I::l:e main steam iNo a T6infalves from inclusionleak in the co B, opkn rate for the local leak rate tests (Section 6.2.6 o SER 6); parayaph exemption from the requirements of paragraph III. f4AppendLx J JK. . B ,

exempting leakage from the valve packing and the y-to-bonnet seal of valve IE51-F374 associated with containment penetration INC-44 from inclusion in the combined leaka e rat for penetrations and valves - op4 BJ et aa_ 'ai la --atfef frM ~the rea"irNaat subecttoTygeBandC-agr:ph ..l.5 C(if te re-Act'Ne th!M Typa a test of ea5h'O N 3 P"'* 50' pIO y::r : rife peri:d d::e sth;eciaT~cp1_an+' i M' -h"t_ M fer_th=

~ances regarding10-y-ana!aateach Q::rtic in::::t_ ten:.

exemption, except for tem /(a): . . above, are identified in the referenced section of the safety eva uation report and the supplements thereto.

(SER wpperb.g A erd ed (,,2 h racM y Opodg License ble. MPF-62}

Amendment No. pd

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i An exemption was previously granted pursuant to 10 CFR 70.24. The 4

exemption was granted with NRC materials license No. SNN-1886, issued November 27, 1985, and relieved IP from the requirement of having a ,

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criticality alara system. IP is hereby exempted from the criticality l

alarm system provision of 10 CFR 70.24 so far as this section applies to the storage of fuel assemblies held under this Itcense.

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-Th :p::1:14reamstances-regarding the ex7 tion-identified-fa-14em-(4}

! -above-are-4 dent 4fied-in-the-safety-evaluation-accompanying fd- xt-

-Nov-62-to-this-Heenser-Th spcial-eircumstenees-regarding S:

-ex7 tie: identified in Ite= (e) deve-are-identiff=d in the nfety-

-evahatier, ac;c anying fn..6.:..t No. 83 -to-this-Meenser-

{ These exemptions are authorized by law, will not present an undue ris

to the public defense and security. health The andexemptions safety, and are consistentgrx ommon acip) i in items (b) ,.. . M above are i i

granted pursuant to 10 CFR 50.12. With these exempt on , lAFfacility will operate, to the extent authorized herein, in conformity with the i application, as amended, the provisions of the Act, and the rules and i

regulations of the Comission. ,

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E. The Itcensees shall' fully implement and maintain in effect all j l i

provisions of the Comission-approved physical security plan, guard  ;

training and qualification, and safeguards. contingency plans including

amendments made pursuant to provisions of the Miscellaneous Amendments -

and Search Requirements revisions to'10 CFR 73.55 (51 FR 27817 and 27822 and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans), which contain Safeguards Information protected under 10 C I

73.21, are entitled: "Clinton Power Station

  • Physical Security Plan,"  ;
  • with revisions submitted through May 27, 1993; "Clinton Power Station i Training and Qualification Plan," with revisions submitted through l l May 27, 1993; and "Clinton Power Station Safeguards Contingency Plan,"

j with revisions submitted through May 27, 1993. Changes made in accordance with 10 CFR 73.55 shall be implemented in acconiance with the

schedule set forth therein.
F. IP shall implement and maintain' in effect all provisions of the approved i

fire protection program as described in the Final Safety Analysis Report as amended, for the Clinton Power Station, Unit No.1, and as approved i in the Safety Evaluation Report (NUREG-0853) dated February 1982 and i

Supplement Nos. I thru 8 thereto subject to the following provision:

IP may make changes to the approved fire protection program j

i without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain

)i safe' shutdown in the event of a fire.

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i Except as otherwise provided in the Technical Specifications or Environmental Protection Plan, IP shall report any violations of the i

requirements contained in Section 2.C of this license in the following j manner: initial notification shall be made within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the NRC Operations Center via the Emergency Notification System with written followup within thirty days in accordance with the procedures described in 10 CFR 50.73(b), (c), and (e).

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Definitions 1.1 1.1 Definitions (continued)

EMERGENCY CORE COOLING The ECCS RESPONSE TIME shall be that time interval SYSTEM (ECCS) RESPONSE from when the monitored parameter exceeds its ECCS TIME initiation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include sitesel generator starting and sequence loading de. lays, where applicable. The response time may 1* be aeasured by means of any series of sequential, overlapping, or total steps'so that the entire respo.1se time is measured.

END OF CYCLE The E00-RPT SYSTEM RESPONSE TIME shall be that RECIRCULATION PUMP TRIP time interval from initial movement of the (E0C-RPT) SYSTEM RESPONSE assoc'.ated turbine stop valve or turbine TIME control valve to complete suppression of the electric arc between the fully open contacts of the recirculation pump circuit breaker. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

ISOLATION SYSTEM The ISOLATION SYSTEM RESPONSE TIME shall be that RESPONSE TIME time interval from when the monitored parameter exceeds its isolation initiation setpoint at the channel sensor until the isolation valves travel to their required positions. The response time may be measured by means of any series of l sequential, overlapping, or total steps so that the entire response time is measured.

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The ==i-" :ll:":ble primary :: tain ::t le:k:ge r:te, L,, : hall be 0.55*' Of pr%.:ry : tai :::t

fr "^ight per d:y :t the ::1 A ted pe:k
t:t :: t pre::"re (P,). ,

(continued)

CLINTON 1.0-3 Amendment No. p/

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The results of SR 3.6.1.1.2 shall be - NOTE -

included when evaluating compliance with 3 SR 3.0.2 is not-this limit. eppl4eable-Perform required visual examinations and In accordance leakage rate testing, except for primary with 40 CFR 50, containment air lock te tin in Append 4x-,)r-as-accordance wit .0 . R 5., nppendix,h- modif4ed by-a: =0..10. -y :pproved-exempt 4ensr approved-

exempt 4ons

-The-l e akage-ra te-accep tance-cr4terdone A Pet ,ary c: 1.0 L,. Howevee r -during-the-f4rst-en4t- f c..,43;,,,a

-sta rtup-foilow ing-te st4ng-performed-in-- I Leage Rak '

-accordance-with 10 CFR 50, Appendix,h-as- J red,3

-modif4cd by cpproved-exempt 4ensy--the-

-leakage-rate-accept-ance criteria-arc Program. l

-<--0.5 L, for-the Type S ar.d Type C test 4r end < 0.75 L, for-the Type A-testy tk Pre ~ary c.,+ain%+ Leaka3e. Ra+e TeShq Prog ram ,

_ _ _ _ _ ,u,nv, ,e- .

/SR 3.0.2 i: net

>applicabic-SR 3.6.1.1.2 Perform leakage rate testing of Primary >In accordance Containment Hydrogen Recombiner System with 10 CFR 50, outside its containment isolation valves lAppendiv J. as at P,. mediried by-

{y approved-exemptions-

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+ 6 Pr;~ 3cy c .,+a;,~a Leakage. Ra4c.

Tes L3

% ram CLINTON 3.6-2 AmendmentNo.f!I

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- - Primary Containme,nt Air Locks , ,, ,

t 3.6.1.2 ; . ,

s SURVEILLANCE REQUIREMENTS' SURVEILLANCE FREQUENCY SR 3.6.1.2.1 ------------------NOTES------------------

1. Only required to be met during MODES I, 2, and 3.

l 2. An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.

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3. Results shall be eval ted against cce tance criteria 3 6.1.1.1 3N\i can b in ::. rdar. e with .0 CTR 50, MOTE-SR 3.0.2 is not I

f;; dix J, :: ::dified by :ppr:ved  ! :ppl10:ble ext ptfeny. _ ._______-_

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! Perform required primary' containment air In ac c  ;

in acco 10 CTR 50, l lock e r te es i with .0 CF.,.0, a p;;d.x ,., :: ::_ified ,.;;;; dix J, ::

ty :ppr ::d ex=;tient. f--difiedby

r:::d The ac- ;tanca criteria for air leck Le;x;--tion -

t:: ting :re:- 4h, p..,marq

n. .. .i i . i ,o i..v. . .+. 4, Conbmed

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~ 2 5'Eih:5+.:E'tE:tE I5t S .5,.

/T sha3

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b. F r :: h d:Or, le:k:ge r:t- i: yeam .

l '" - " " -- th^ g:; S tr:: th- ,

l d::r :::1: i: pre: erized te i P,. J l the Pr mary CoaWe4 i eakass RA l

-fest*m , Pe. cram /

SR 3.6.1.2.2 ------------------NOTE-------------------

l Only required to be performed upon entry l or exit through the primary containment l

l air lock.

Verify only one door in the primary 184 days containment air lock can be opened at a time.

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3.6-8 Amendment No. Jif CLINTON

! 9 PCIVs 3.6.1.3 l l

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') SURVEILLANCE REQUIREMENTS (continued) i l I l

l SURVEILLANCE FREQUENCY l i

i SR 3.6.1.3.4 Verify the isolation time of each power In accordance ,

i operated and each automatic PCIV, except with the '

MSIVs, is within limits. Inservice ,

Testing Program '

l SR 3.6.1.3.5 ------------------NOTE---------- - --- l Only required to be met in MODES 1, 2, and 3.

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

Perform leakage rate testing for each Once within 92 l primary containment purge valve with days after resilient seals. opening the valve AND

~) F "0TE

'SR ?.0.2 i: .0t

2pplicable i

In acco danc with;10 Cf.,50,

(?.ppend!Y 2, 2t

-^d!#ied by t6. Pr.mey Coela6.,d Lea 63 Ne Te 6,3 Pegram  ;"" {

SR 3.6.1.3.6 Verify the isolation time of.each MSIV In accordance is 1 3 seconds and s 5 seconds. with the Inservice Testing Program SR 3.6.1.3.7 Verify each automatic PCIV actuates to 18 months the isolation position on an actual or

simuisted isolation signal.

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(continued) 1 CLINTON 3.6-17 Amendment No. B5 #I

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PCIVs
3.6.1.3 SURVEILLANCE, REQUIREMENTS (continued) .

SURVEILLANCE FREQUENCY SR 3.6.1.3.8 s-----------------NOTE------------------ -- -NOTC Only required to be met in' H00ES 1, 2, SR 3.0.2 i:

and 3. r.ct :pplic:ble Yerify the combined leakage rate .for all In acc c secondary containment bypass leakage witht10 era S0, paths is s 0.08 L, when pressurized to ,; App:: dix J.,  ::-

dP.a redif! M Sy jappreved

_su.

Ac Aary ConlaWme.d Ledage Rake Tesb3 %re *^'~r""-

SR 3.6.1.3.9 ------------------NOTE------------------ ----NOTC -- -

Only required to be met in MODES 1, 2 SR 3.0.2 i:

and 3. r. t :ppl1 :ble -

Verify leakage rate through each main In accordanc steam line is s 28 scfh when tested at with/10 CTR S0, a P,. App:: dix 2, :q- l

, medi#!ed by '

(:ppr:ved

... >s...

~'~~"~""~

tbc. Pet. nary C faiW.hLea.3y. Rde Tesb 3 Pega-

+_ = _-_ _ - - _ _ ~

SR 3.6.1.3.10 ------------------NOTE------------------ ----NOTC Only required to be met in MODES 1, 2, SR 3.0.2 i:

and 3.  : t :pplic:ble-

_ Verify _combinedleakaaerateh Inaccgne Q:-- M Oti:llaf b^r effCIV:, rough withA1. ...,50, ydrostaticalTy~ te~sted~ lines M at App 00 dix J, ::

penetrate the_ primary contai_nment is net-  ::dified by

~; isiinTG,t ~tEERWitEdV:Tv~cT re :pproved i :ted at _ 1.1 P -a WilW limit 5 - OX25 L__?tiO2 w

--- .,,,c,,. " _ . , ,

he Pc;macy Con 4ainmed Leakay Rete hb3 (continued)

CLINTON 3.6-10 Amendment No.,85

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  • Programs and Manuals .

5.5 5.5 Programs and Manuals (continued) 5.5.11 Technical Specifications (TS) Bases Control Procram This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative centrols and reviews.
b. Licensees may make changes to Bases without prior NRC approval 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 USAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59.
c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the USAR.
d. Proposed changes that meet the criteria of either Specification 5.5.11.b.1 or Specification 5.5.11.b.2 above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

5.5.12 Ultimate Heat Sink (UHS) Erosion. Sediment Monitorina, and i Dredaina Procram A program to provide mitintenance on the UHS in the event -

inspections of the UHS dam, its abutments, or the UHS shoreline indicate erosion or local instability. This program shall ensure that the UHS is maintained in such a way as to achieve the following objectives:

a. During normal operation, there will be a volume of water in the UHS below elevation 675 sufficient to receive the sediment load from a once-in-25-year flood event; and
b. Still be adequate to maintain the plant in a safe-shutdown ND condition for 30 days under meteorological conditions of the severity suggested by Regulatory Guide 1.27.

1 cAany. b sqe. 6c.

CLINTON 5.0-16 AmendmentNo.JHI

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. Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.13 Primary Containment Leakaae Rate Testina Proaram 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 modified by approved exemptions. This program shall be in accordance with the i guidelines contained in Regulatory Guide 1.163, " Performance-Based I Containment Leak-Test Progra", dated September 1995, except that Bechtel Topical Report BN-TOP-i is also an acceptable option for performance of Type A tests.

The peak calculated containment internal pressure for the design basis loss of coolant accident, P,, is 9.0 psig.

The maximum allowable primary containment leakage rate, L,, at P,,

shall be 0.65% of primary containment air weight per day Leakage Rate acceptance criteria are:

a. Primary containment leakage rate acceptance criterion is s 1.0 L,. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are s 0.60 L for the Type B and Type C tests and s0.75L,forTypeAfests;
b. Air lock testing acceptance criteria are:
1) Overall air lock leakage rate is s 5 scfh when tested at a P,,
2) For each door, leakage rate is s 5 scfh when the gap between the door seals is pressurized to a P,.

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

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

CLINTON 5.0-16a Amendment No.

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! Attachment 4 to U-602565 LS-95-014 l Page 1 of15 i

l Technical Specification Bases Changes I

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.t SR Applicability B 3.0 BASES 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 Therdore,wken a being performed is the verification of conformance with the L

tes+ ;.d.rval ts sed;4 SRs. The exceptions to SR 3.0.2 are those Surveillances for L in Ac resvtationy se which the 25% extension of the interval specified in the Frequency does not apply. These exceptions e_ sta_ted i ksi interi<al ca"ad

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the individual Specifications.f An en=p.e of 95:Ye S. 3.0.2-eMby A TS s .NyMa-SurveH44 Wee +Mh : Frequency cf "in a.J A Ts will + ken 10 CFR " ^ refjffeLhL-ineLJe a eJoTe stat -- rc"^' exempt 4

-accordance-with # ;;664M utrements aTfMT'endipJr-4;rregul "M 3.0,2 is ,J e over the TS.3 e .ves-ex T:irrtest~Tnterial-spee4f4ed-4n-the-regulat-lensr 3M**bo^^*Y Thereforer-there-1s-a-4fote-4n-the-Frequency-statingr d * * ""*P6 de" 2SR-370.2 is-not-app 14eable?

+Le. +cshJeeul is n a spe d ej iniLe As stated in SR 3.0.2, the 25% extension also does not apply to the initial portion of a periodic Completion Time that

- D43 3 g requires performance on a "once per..." basis. The 25%

h M'.'" b ~^"Y extension applies to each performance after the initial l M 8'~" 4 Lea % e performance. The initial performance of the Required Rafe.Tedq Resam, Action, whether it is a particular Surveillance or somo "sR 3.o.2. is no+ other remedial action, is considered a single action with a appeak," This single Completion Time. One reason for not allowing the 25%

W *9i S Prog extension to this Completion Time is that such an action usually ve'ifies r that no loss of function has occurred by 6mne h pecT**' checking the status of redundant or diverse components or ald incJudes y accomplishes the function of the inoperable equipment in an cAens u cR fest alternative manner.

in4ecels .

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 specified limits when a Surveillance has not been completed within the specified Frequency. A delay period of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is less, applies from the point in time that it is discovered that the Surveillance has not been performed in accordance with SR 3.0.2, and not at the time (continued)

CLINTON B 3.0-12 RevisionNo./

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Primary Containment B 3.6..l.1 BASES BACKGROUND e. The leakage control system associated with the main (continued) steam lines is OPERABLE, except as provided in LC0 3.6.1.8. " Main Steam Isolation Valve (MSIV) l Leakage Control System (LCS)"; and

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f. The priniary containment leakage rates are within the limits of this LCO. 4 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.

- 1 APPLICABLE The safety design basis for the primary containment is that SAFETY ANALYSES 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 wit.hin primary containment is a LOCA. In the analysis of this accident, it is assumed that primary containment is .0PERABLE'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 containment are presented in References 1 and 2. The safety analy'ses assume a nonmechanistic fission product release following a DBA, which forms the basis for determination of {

offsite doses. The fission product release is, in turn, j based on an assumed leakage rat from the primary conteinment. 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 (La ) is 0.65% by weight of the containment and desgn b3St5 M drywell air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at theMmaximum peak containment pressure (P,) of 9.0 psig (Ref 4).

Primary containment satisfies Criterion 3 of the NRC Policy Statement.

(continued)

CLINTON B 3.6-2 Revision No. /

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- Primary Containment B 3.6.1.1 BASES (continued)

LCO Primary containment OPERABILITY is maintained by limiting leakage to s 1.0 L,, eEciptJrio_ry st startup afte PrimaryCoab, eat performinge athirenuireoC0 Cf". "50, ts;=m 2, leakage At th s ~dd .b,;,cd Type ud C ' leakage.must be test.d rre Leak ge Ra4e T*SbS N "#'" onh ance th otkk k gd{,s configuration, including equipment hatches, that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analysis. Individual leakage rates specified for the primary 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 other operational conditions, events which could cause a release of radioactive material to primary containment are mitigated by secondary 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 is not required to be OPERABLE in MODES 4 and 5 to prevent leakage of radioactive i material from primary containment.

ACTIONS A.1 In the event that primary containment is inoperable, primary containment must be restored to OPERABLE status within I hour. The I hour Completion Time provides a period of time to correct the, problem that-is commensurate with the importance of maintaining primary containment OPERABILITY during MODES 1, 2, and 3. This time period also ensures that the probability of an accident (requiring primary containment OPERABILITY) occurring during periods where primary containment is inoperable is minimal.

ACTIONS B.1 and B.2 If primary containment cannot be restored to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant u st be brought to at least (continued',

J CLINTON B 3.6-3 Revision No. p r l

l Primary Containment B 3.6.1.1 BASES'

}

ACTIONS B.1 and B.2 (continued) ,

I MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The '

i allowed Completion Times are reasonable, based on operating experience, to reach the required plint conditions from full

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power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.1.1.1 REQUIREMENTS

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coa b .=< t Maintaining the primary containment OPERABLE requires thePA.,[rT .) compliance with the visual examiaalisns an rate test reau_irement  ::-d .x J $^. . _3),_ ::

Leaka3e Raie W 3 b %'*m - 0-OM0'd3y_ ipp yf ^V^d tifc"CFP3f, 2;^^ a^ failurie'to meet air Tock leaYage testing (SR 3~.6.1.2.1), secondary containment bypass leakage (SR 3.6.1.3.8), resilient seal primary containment N R wa,y c.e4a;, d purge valve leakage testing (SR 3.6.1.3.5), main steam g' , isolation valve leakage (SR 3.6.1.3.9), or hydrostatically g g1 tested valve' leakage .(SR 3.6.1.3.10) does not necessarily hgeam result in a failure of this SR. The impact of the failure to meet these SRs must be eval the Type A.

- - - - C "'; ' . . ,

nce cri tp.. -.,-_ft.,--,l. .. .. s lea acyrio_r - ,

t startup after pe orming a re ed M J R 50 ?

.4 u leakage test is requir to b .6 i comb Type B and C leakage, an .75 , or overall Type A leakage. At all other tim etween required leaksge rate tests, the acceptance criteria is based on an overall Type A leakage limit of s 1.0 L,. At s 1.0 L, the offsite dose consequences are bounded by the assumptions f t.e ^

eauenc bym Un . , ,

t ufatr

.ppeanalysj st._ g In%g o r _yj;y_is_ require _-gg .x-_7te ,:, 13;;,

AeRi.,aey c.,6 gg 3, ,5

'^
jd:::: top?ly.

kaki,3e Rak hb3 SR 3.0.2 @hjch all

'" This Surveillance is modified by a Note that requires the leakage rate results of SR 3.6.1.1.2 for the Primary Containment Hydrogen Recombiner System (each loop) to be included in determining compliance with required limits.

This can be accomplished either by having the loops in service during the ILRT, or if the loop is not in service during the ILRT, by separately measuring the leakage and including it in the measured ILRT results.

I (continued) i i

l 1

CLINTON B 3.6-4 Revision No. S T l

l - . . - . _ - - . . - .

. i*

4 Primary Centainment B 3.6.1.1 BASES J

1 SURVElt. LANCE SR 3.6.1.1.2 .

REQUIREMENTS (continued) With respect to primary containment integrated leakage rate testing, the primary containment hydrogen recombiners (located outside the primary containment) are considered

( extensions of the primary containment boundary. This j

- bc Ecimori requires the smaller of the leakage from the PCIVs that c.43;nmen+ isolate the primary containment hydrogen recombiner, or from g the piping boundary outside containment, to be i luded i g

hkT/," Ef Ts%g Pq,ca m (3 =iud _t.IEF. Ell 5YEbEqE5nE[EEtE5:5E5:)'55EE'n:t5hhlh.

~ _ _ w.~ - _

REFERENCES 1. USAP, Section 6.2.

I 2. USAR, Section 15.6.5. ,

3. 10 CFR.50, Appendix l
4. USAR, Section 6.2.1. l x -

'). p m

t4E t 94-c>t ; Revision 0, '% Ass 4ry Goidelia e. for 5.

l~pt.-e..h3 hefor-a=-Based opb .T 10 cFR Par + 503 APP

  • J < J . "

b

[o. AMS /AM S - S G . B- l*>94; " Amacican t4abal

%4rd 9.r- C.Jairmed Syh Leakay

(

Tedt,,3 Repicemed . "

l

.)

CLINTON B 3.6-5 Revision No M

5 Primary Containment Air Locks i

B 3.6.1.2 l

BASES l

ACTIONS E.1. E.2 and E.3 (continued) positios Also, if applicable, action must be immediately initit>J to suspend OPDRVs to minimize the probability of a l vessel draindown and subsequent potential for fission

~

product rele~ase. Action must continue until 0PORVs'are l suspended.

The Required Actions of Condition E are modified by a Note indicating that LCO 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.

Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.

m mm SURVEILLANCE SR 3.6.1.2.1 h b aty M aaent Leaka,p M e Teska3 hayam

' ^ " -

i REQUIREMENTS Maintaining primary containment air locks OPERABLE requires

. ,n W'&

h k : b 1 t k k[! En~

fN U Y O D M I N TdIthe

) Teakage rate testing requirements with regard to air lock l

leakage (Type B leakage tests). The acceptance criteria were established during initial air lock and. primary containment OPERABILITY testing. The periodic testing l  %

requirements verify that the air lock leakage does not 4* p,.**' g"*'p exceed the allowed fraction of the overall primary required by i nment_teakage rate _._ Thef re_q LeAty M[T*$b3 fl.cgngO6, ~^?pshdiv J,Y iddi#fe'd yappre'fei 4*Mptiens.

lThuc, SP ?.0.2 (.:hich ;11;ws Trcq= ncy extensien:) d::: not gm, i epp47 The SR has been modified by three Notes. Note 1 provides an exception to the specific leakage requirements for the primary containment air locks in other. than MODES 1, 2, and

3. When not operating in MODES 1, 2, or 3, primary I containment pressure is not expected to significantly l increase above normal, and therefore specific testing at elevated pressure is not required. Note 2 states that an inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.

l This is considered reasonable since either air lock door is (continued)

)

CLINTON 8 3.6-13 Revision No..P T l

.s*

Primary Containment Air Locks l

B 3.6.1.2 l

, i.e. , +be wep eree_ cr3erm spec 4ed *n Ac l

I BASES *~*'Y C'"*"' '** '3' *

  • 3 3"

h l "

l SURVEILLANCE SR 3.6.1.2.1 (continued) l REQUIREMENTS capable of providing a fission product barrier in the event of a DBA. Note 3 has been added to this SR, requiring t e ,

l ..

results to be eva ted a ainst the acceptance cr' b iaf9f R PckdC

]

SR 3.6.1.1.14 ' ^ sure: , , ,

glockleakage o b'e i incLeleI 'p-ff T E SS T P err.n ning the overall primary l

con ainmenfleakage rate. -

con b a<e4o Ae Pet ~ wy C.4 ai n~e.,4 t.eaka e Rak Tedi.sg Pro 3 ram rey.res The air ic.ck interlock mechanism is designed to prevent s!multanecus 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 pressure (Ref. 4), closure of either door will l  ;

support primary containment OPERABILITY. Thus, the  !

interlock feature supports primary containment OPERABILITY l white 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 1 designed ar;d that simultaneous inner and outer door opening -

will not inadvertently occur. Due to the nature of this interlock, and given that the interlock mechanism is only challenged when the primary containment air lock door is i

opened, thic, test is only required to be performed upon l entering or exiting a primary containment air lock, but 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.

REFERENCES 1. USAR, Section 3.8.

2. 10 CFR 50, Appendix
3. USAR, Section 6.2.1.
4. USAR, Section 15.7.4.

l l

l CLINTON B 3.6-14 Revision No. K i

.i' PCIVs

. B 3.6.1.3 BASES ACTIONS 0.1. D.2, and 0.3 (continued) closed (refer to the requirements of SR 3.6.1.3.1; if this requirement is not met, entry into Condition A and B, as appropriate, would also b'e required), so that a gross breach

- of primary containment does not exist.

' In accordance with Required Action D.2, this penetration flow path must be verified to be isolated on a periodic basis. The periodic verification is necessary to ensure that primary containment penetrations required to be isolated following an accident, which are no longer capable of being automatically isolated, will be isolated should an event occur. This Required Action does not require any testing or valve manipulation. Rather, it involves verification that those isolation devices outside primary containment and potentially capable of being mispositioned are in the correct position. For the isolation devices inside primary containment, the tim: period specified as

" prior to entering MODE 2 or 3, from MODE 4 if not performed within the previous 92 days" is based on engineering i

judgment and is considered reasonable in view of administrative controls that will ensure that isolation

) device misalignment is an unlikely possibility.

For a primary containment purge valve with a resilient seal that is isolated in accordance with Required Action 0.1, j

SR 3.6.1.3.5 must be performed at least once every 92 days.

This provides assurance that degradation of the resilient seal is detected and confirms that the leakage rate of the gp g.,j primary containment purge valve does not increase during the 7

i Leary M time the penetration is isolated. The normal Fre ue c fr J l l Tes 43 rogram.

P R 3.6.1.3.5 is as re,quireMyMT[s,56,*@i;M

3) , n cd t 10, ); ;g /cd mtte3 - ince more re af

~

tance l

1s p acTd'on a single vaMhIle in this Condition, it is

~

l Therefore, a prudent to perform the SR more often.

Frequency of once per 92 days was chosen and has been shown acceptable based on operating experienc'e.

E.1 and E.2 l

If any Required Action and associated Completion Time cannot be met 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 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed (continued)

)

CLINTON B 3.6-21 RevisionNo./

l

,s*

, PCIVs B 3.6.1.3 l

) BASES

! SURVEILLANCE SR 3.6.1.3.4 (continued)

REQUIREMENTS in a time period less than or equal to that assumed'in the safety analysis. The isolation time and Frequency of this i

SR are in accordance with the Inservice Testing Program.

l I m Q p,. SR 3.6.1.3.5 C"d * '* * ^i for primary containment purge valves with resilient seals, ond the test requirements

( Leakage. Rale. addi oft,

- ~ 4 gin is required to ensure l eseng gam )

j OPE N p M ._.. .,

W cn erion for this test is Since cycling s 0.01 L, when pressurized to Pa, 9.0 psig.

these valves. may introduce additional seal degradation ,

(beyond that which occurs to a valve that has not been opened), this SR must be performed within 92 days after opening the valve. However, operating experience has demonstrated that if a valve with a resilient seal is not stroked during an operating cycle, significant increased l leakage through the valve is not observed. Based on_this I

ENM *SSSb!0,"S*5 b* b8N '

-) Codamment 7 " g $hed'.7 N 555E5E55 = U5 5NfSb[5hp55.5I d 5.'

l

, Leak ge & 3,c ,17,;g:11 g p7eg y exte : ten:) ge : :t :pp!~,

Tch 3 gram The SR is modified by a Note stating that the primary containment purge valves are only required to meet leakage ,

rate testing requirements in MODES 1, 2, and 3. If a LOCA inside primary containment occurs in these MODES, purge valve leakage must be minimized to ensure offsite radiological release is within limits. At other times when the purge valves are required to be capable of closing i

(e.g., during handling of .rradiated fuel), pressurization concerns are not present and the purge valves are not required to meet any specific leakage criteria.

SR 3.6.1.3.6 Verifying that the full closure isolation time of each MSIV is within the specified limits is required to demonstrate OPERABILITY. The full closure isolation time test ensures that the MSIV will isolate in a time period that does not exceed the times assumed in the DBA analyses. The Frequency of this SR is in accordance with the Inservice Testing Program.

(continued) l I B 3.6-25 Revision No. g

! CLINTON l

l

..%*

  • 1 1

\ .

(~*

B 3.6.1.3 l

BASES

? .

SURVEILLANCE SR 3.6.1.3.7 REQUIREMENTS (continued) Automatic PCIVs close on a primary containment isolation l

signal to prevent _ leakage of radioactive material from l

primary containment following a DBA. This SR ensures that each automatic PCIV will actuate to its isolation position l on a primary containment isolation signal. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.1.7 overlaps this SR to provide l

complete testing of the safety function. The 18 month Frequency is based on the need to perform this Surveillance under the conditions 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 Frequen;;y was concluded to be acceptable from a reliability standpoint.

i l SR 3.6.1.3.8 This SR ensures that the leakage rate of secondary containment bypass leakage paths is less than the specified

. leakage rate. This provides assurance that the assumptions .2 in the radiologicul evaluations of References 1, 2, and 3 l

are met. The leakage' rate of each bypass leakage path ie assumed to be the maximum pathway leakage (leakage through l

the worse of the two isolation valves) unless the penetration is isolated by use of one closed and de-activated automatic valve, closed manual valve, or 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 isolation l

valves in the penetration are closed, the actual leakage l rate is the lesser leakage rate of the two valves. This method of quantif" n mum pa a is o to be usedMhis Sp,[i .e. , ^.ppenax 3 m4 mum-j>athway-4ea -age- l th, d e ] NN$INES5S7

^

h+

%k $ egam.P fTiIisSR simp g"; Frc=0r~ adcMacc.denp y imposes ext ^reptance Wdee':eria. pc.t App' Secondary containment bypass leakage is considered part of La -

A Note is added to this SR which states that these valves are only required to meet this leakage limit in MODES 1, 2 and 3. In the other conditions, the Reactor Coolant System I

(continued)

CLINTON B 3.6-26 Revision No. JA' l

PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.8 (continued)

REQUIREMENTS is not pressurized and specific primary containment leakage limits are not required.

SR 3.6.1.3.9 The analyses in References 1, 2, and 3 are based cn leakage that is less than the specified leakage rate. Leakage through each main steamline must be :s; 28 scfh when tested at P, (9.0 psig). The MSIV leakage rate must be verified to 1,0 in accordance with the assumptions of References 1, 2, and

3. A Note is added to this SR which states that these valves are only required to meet this leakage limit in MODES 1, 2, and 3. In the other conditions, the Reactor Coolant System is not pressurized and primary containment leakage a a R re Leakay We exemptions;-thus,-SR-3A2-(which-al-low: Frequency-Tes43pgam . extensions}-does-not-apply, __

(,,f 1 3 p , vi ~ s A e. W ev 4 er J P C W 5 SR 3.6.1.3.10 A, tesW >& h i.t P, )

Surveillance of hydrostatically tested lines provides assurance that the calculation assumptions of Reference 4 are met. The combined leakage ratesraust be demonstrated at Ae hary ne g uency of the le Ka e test re_quirements of 4 cc.,4 aumed Re erence T : _ 4fied y-approved-ex:5;.ica:;-t s -SR- r Lea g eae GA2-(which-allows fre_q : :y ext: : ices) ( pg

  • D '"* This SR is modified by a Note that states that these valves are only required to meet the combined leakage rate in MODES 1, 2, and 3 since this is when the Reactor Coolant System is pressurized and primary containment is required.

In some instances, the valves are required to be capable of automatically closing during MODES other than MODES 1, 2, and 3. However, specific leakage limits are not applicable in these other MODES or conditions.

SR 3.6.1.3.11 This SR requires a demonstration that each instrumentation line excess flow check valve (EFCV) is OPERABLE by verifying that the valve activates within the required differential (continued)

CLINTON B 3.6-27 RevisionNo./

l I

pCIVs B 3.6.1.3 BASES ,

t SURVEILLANCE SR 3.6.1.3.11 (continued)

REQUIREMENTS pressure range. This SR provides assurance that the instrumentation line EFCVs will perform so that predicted radiological consequences will not be exceeded during the

! postulated instrument line break events (Ref. 7). The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

REFERENCES 1. USAR, Chapter 15.6.5.

2. USAR, Section 15.6.4.
3. USAR, Section 15.7.4.

f 4. USAR, Section 6.2.

l

5. USAR, Table 6.2-47.
6. 10 CFR 50, Appendix , 0;dio^ E5 l
7. Regulatory Guide 1.11.

l l

l s

CLINTON B 3.6-28 Revision No.)Y'

l . ..

'^ Drywell B 3.6.5.1 l

BASES SURVEILLANCE SR 3.6.5.1.1 (continued)

REQUIREMENTS properly accounted for in the measured bypass leakage and that each air lock is tested periodically. The leakage test is performed every 18 months, consistent with the difficulty of performing-the test, risk of high radiation exposure, and the remote possibility that a component failure that is not identified by some other drywell or primary containment SR l

l might occur. Operating experience has shown that these l components usually pass the Surveillance whe.n performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

l In addition, if two consecutive tests fail to meet the leakage limit, a test shall be performed at least every 9 months until two consecutive tests meet the limit, at which  ;

time the 18 month Frequency may be resumed.

SR 3.6.5.1.2 l l

l The exposed accessible drywell interior and exterior surfaces are inspected to ensure there are no apparent physical defects that would prevent the drywell from I

) performing its intended function. This SR ensures that l

i drywell structural integrity is maintained. The Frequency was chosen so that the interior and exterior surfaces of the drywell can be inspected in conjunction with the inspections )

l of the primary containment required by 10 CFR 50, Appendix J  !

(Ref.2). Due to the passive nature of the drywell

[ structure, the specified Frequency is sufficient to identify l

component degradation that may affect drywell structural ,

integrity.

REFERENCES 1. USAR, Chapter 6 and Chapter 15.

2. 10 CFR 50, Appendix , Opho 15,.

1 i

l

. i

I .

l CLINTON B 3.6-105 Revision No. JA I

. -- . .. . ~ . . . ..

s *.'

Orywell Air Lock B 3.6.5.2 I BASES SURVEILLANCE SR 3.6.5.2.2 (continued) l REQUIREMENTS The Surveillance is modified by a Note requiring the Surveillance to be performed only upon entry into the drywell.

l SR 3.6.5.2.3 l

This SR requires a test to be performed to verify overall air lock leakage of the drywell air lock at pressures

e 3.0.psig. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply l during a plant outage and the potential for violating the drywell boundary. 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 L

acceptable from a reliability standpoint.  ;

i This SR has been modified by two Notes. The first Note indicates that an inoperable air lock door does not invalidate the previous successful performar. e of an 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.

l The Surveillance is modified by a second Note requiring the g

air lock to be pressurized to 19.7 psid prior to performance of the overall air lock leakage test. The 19.7 psid l

I differential pressure is the assumed peak drywell pressure i l

expected from the accident analysis. Since the drywell l l

pressure rapidly returns to a steady state maximum

-- differential pressure of 3.0 psid (due to suppression pool

-- vent clearing), the leakage is allowed to be measured at this pressure. ,

l l

REFERENCES 1. 10 CFR 50, Appendix 3 Op b E.

2. USAR, Chapters 6 and 15.

CLINTON 8 3.6-112 Revision No. /

' D, Attachment 5 to U-602565 LS-95-014 i

Page1of2 l

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1 1

l I

t Clinton Power Station l 10CFR50, Appendix J - Option B, Implementation Schedule 3

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4 i

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I L

t n

(

)

e 1

f I

(

, e.

LS-95-014 Page 2 of 2 CLINTON POWER STATION 10CFR50, APPENDIX J - OPTION B IMPLEMENTATION SCIIEDULE l 5/1/96 Commence development of document which will define and control the l

Primary Containment Leakage Rate Testing Program.

l l

This may include:

1. Description ofProgram Requireme its.
2. Control of Type A, B, and C testing intervals.
3. Methods of evaluating test results.
4. Administrative Limits for Type B and C components.

l 5. Maximum and Minimum pathway leakage algorithms for Type C components.

6. Record keeping requirements.

l 7. Database specification.

8/1/96 Complete development of Primary Containment Leakage Rate Testing i Program document.

l 10/1/96 Completo changes to testing procedures necessary to support implementation of 10CFR50, Appendix J - Option B, and the Primary Containmem Leakage Rate Testing Program.

I Complete initial frequency determination and justification for extending testing intervals of components which will not be tested during the sixth l refueling outage. Type B and C test frequencies will be based on the l most recent Local Leak Rate tests performed.

t 10/13/96 Implementation date of 10CFR50, Appendix J - Option B, Primary (Start ofRF-6) Leakage Rate Testing Program.

Note: It is intended that the implementation date of the 10CFR50, l Appendix J - Option B, Primary Containment Leakage Rate Testing l Program coincide with the actual starting date of the sixth refueling outage, currently scheduled for 10/13/96. Should operational requirements and events alter start of the refueling outage, then the actual date ofimplementation will shift accordingly.

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