Text

Forwards Responses to NRC 980707 RAIs Re Proposed Improved Tech Specs Submitted by Util on 970527
	ML20236M874
Person / Time
Site:	McGuire, Mcguire
Issue date:	07/07/1998
From:	Tuckman M; DUKE POWER CO.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	TAC-M98964, TAC-M98965, NUDOCS 9807140299
	Download: ML20236M874 (190)
	v • d • e

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Duka Pown Company A 1h,h, Ewy Camp,my

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EC07H A f** % % 526 South Church Street EO. Box 1006 Charlotte, NC 28201 1006 M. S. Tudman Executive Vice hesiden, (M)382-22% om Nudear Generation (M)3824360 m l

July 7, 1998 U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Document Control Desk

Subject:

McGuire Nuclear Station Docket Numbers 50-369 and -370 Improved Technical Specifications, Supplement 6 TAC Nos. M98964 and M98965 By letter dated July 7, 1998, the NRC transmitted requests for additional information (RAI) related to Section 3.3 and 5.0 of the proposed Improved Technical Specifications (ITS) submitted by Duke Energy May 27, 1997. Enclosure 1 of this letter provides responses to the RAI on Section 3.3 and 5.0.

Duke Energy's meeting with the NRC on June 17 and 18, 1998 resulted in open items associated with previous RAI responses on Sections 3.5, 3.6, and 3.9. Resolution of these open items are included as Enclosures 2, 3, and 4.

Enclosure 5 incorporates corrections or clarifications identified by DEC internal reviews.

Immediately following each staff comment or DEC identified additional item are changes to the ITS submittal necessary.

to resolve the item and supplement the initial ITS submittal. Changes are denoted by revision bars to facilitate staff review.

The pages provided in Enclosures 1, 2, 3, 4, and 5 in this response replace the corresponding pages in the May 27, 1997 submittal, Supplement 1 provided March 9, 1998, Supplement 2 provided March 20, 1998, Supplement 3 provided April 20, 1998, Supplement 4 provided June 3, 1998 and Supplement 5 dated June 24, 1998. 7

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l U. S. Nuclear Regulatory Commission July 7, 1998 Page 2 Changes associated with the above comments encompass

. clarifications in the Bases for each section, additional information requested by the Staff in the Discussion of Changes,:and additional clarifications identified by DEC fnternal reviews. All changes have been' determined to be within the scope of the' original PORC and NSRB' reviews.

Pursuant to 10.CFR 50.91(b)(1), a copy of this amendment has been,provided-to.the appropriate State of North Carolina officials.

If any additional information is'needed, please call Lee A.-

Keller.at.704-382-5826.

Very truly ygurs, f5;.;G-4:=~

M. S. Tuckman-Enclosures i

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___________-______________ _ I

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U. S. Nuclear Regulatory Commission July 7, 1998 Page 3 xc: w/ enclosures Mr. L. A. Reyes Administrator, Region II :

U. S. Nuclear Regulatory Commission '

Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, GA 30303 Mr. F. Rinaldi U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, DC 20555 Mr. S. M. Shaeffer Senior Resident Inspector McGuire Nuclear Station Mr. Mel Frye Division of Radiation Protection i 3825 Barrett Drive '

Raleigh, NC 27609-7221 l

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4-O ENCLOSURE 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3 AND 5.0 0

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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . . . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ . _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ ___._ ___ ___ ___ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ . _ __ ______________.J

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( 3.3.3, PAM Instrumentation

) ]

3.3.3-06 LCO 3.3.3 '

JFD 10 Justification JFD 10 states that LCO 3.3.3 actions have been revised to support the plant j specific design which includes single channel PAM variables. The result is to reformat the NUREG LCO 3.3.3 to take advantage of the less restrictive actions. While this approach was approved for the Vogtle ITS and is proposed for the Byron /Braidwood plants, and McGuire units, the staff notes that NUREG TS were adopted or are proposed with some minor i variations for Farley, Robinson, Diablo Canyon, Wolf Creek, Comanche Peak, South Texas l

and Ginna. Where required, the staff can accommodate small changes in the LCO 3.3.3 i format to address unique required actions for single channel and diverse variable functions as !

appropriate. Comment: Provide rev! sed CTS and STS markups with supporting DOCS and JFDs.

DEC Response:

DEC disagrees that any change to the ITS is necessary. The issues identified by the staff are of a presentation nature and are not technical. For Farley, there are not any single channel functions and modifications to the STS format were not necessary. A review of the Robinson TS indicates a signifcant variation to the STS format to accomodate the single channel functions. For McGuire, reliance on the diverse variables was not utilized for single channel 7N functions because none are specified in the UFSAR for those functions. The inclusion of

) single channels within a unit TS was determined acceptable by the NRC during the review for RG 1.97 based on diversity of indication. DEC has proposed a TS that meets the current licensing basis and provides the desired operational flexibility based on channel diversity for a small number of functions. The ITS was modeled after the only available models at the time of development which were Vogtle and the original South Texas submittal. The other plants referenced in the staff comment were submitted on or about the same time as the DEC submittal. In any case, it is evident that the STS does not function for variables with single channels and that utilities with this design must modify the STS to incorporate the approved licensing basis. DEC believes that the proposed ITS meets all CTS requirements and is consistent with the licensing basis for diverse variables and should be acceptable based on staff approvals of similar ITS. The proposed ITS does minimize the changes to the STS format, however, additional actions are necessary to accomodate the diverse channel aspects.

The only substative difference between the Robinson TS and the proposed ITS is that Robinson directs which conditions are applicable by the use of Notes and the DEC model !

uses the Table to direct entry into required actions consistent with 3.3.1 and 3.3.2. This difference is in presentation only and not in technical content. Operators have been trained at both sites using this format and a change in presentation would result in a significant impact without any safety benefit. Additionally, DEC has had discussions with the NRC technical branch which reviewed the proposed PAM changes and were told that the changes l were acceptable.

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mc3_cr_3.3a 0 July 1,1998 O

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McGuire & Catawba improved TS Review Comments Section 3.3 instrumentation i

3.3.3-07 ITS SR 3.3.3.1 Channel check requirements are changed to apply only to those functions normally energized.

Comment: Provide a DOC for the substantive changes proposed for ITS channel check

. surveillance.

DEC Response:

DOC L32 is added to justify this change.

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l mc3_cr_3.3a .1 July 1,1998 9

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[ 3.~$ INSTRUMENTATION Q -

43.3 @ ACCIDENT MONITORING INSTRUMENTATION

@ Tam wuun ION FORAPERAIHm Q

LCo 3.3.3@ The accinaar -nring instrumentation Tchaonfisenowlp in he ts.dM)

Table shall be OPERABLE.

A.l APPLICAsl IIf: MODES 1, 2, and 3. M" p g, g JN i* '

m,- ACTION * -

's usentrT Q With number of #PtRAutt acciae is less th V the Recuired N nitoring r of Channe shown i nt M

I \cha Ta 3.310JrestorVthe inoperablychannel(s) 'to DPERA8LE stf f ul 1,Acw 3g with l

w c. 12 h rs . f days. or h(in at least W6T SHUTDOWN wit)fn the nexy J

. With the n of OPERABLE accident nt toring 1r.am ..t. ion channels les .than the Minimum Chan is OPERABLE requi s of Table 3.3-1 , restore the inoperab channe'(s)toOPERA status. ,1 t.19 according o b.1 and b.2 below: --

- IWST b.1 Instruments 1-15: within hourst or be in at least HOT 4CV*#

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[SHUTDOWNwithinthenext12 hours. J In b.2[1,struments 16' anni 17e /accordino to T*chnica, specific aa -

,$g The provisions of Specification 3.0.4 are not applicable.

NectMwH M SURVEILLANCE REOGLREMENTS AW W 3 l'I N Each accident monitoring instrumentation channel shall be

E demonstrated OPERABLE by performance of the CHANNEL CHEC nd CHANNEL 5" M- CALIBRATION operations at the frequencies shown in Table .3-7.

or M Er instrupMen bcLAr4f '

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3/4.6.4 BUSTIBLE GAS CONTR90

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lHY MONITORS I

U.IMITING CONDITION F OPERATION /

v)CLuceo N TAtt . . Two independent containment hydrogen monitors shall be OPERABLE.

\.V5- 1 APPLICABILITY: MODES 31 @

ACTION:

. ' ACT1*d S With one hydrogen monitor inoperable, restore the inoperable monitor to OPERABLE nours e status within 30 days Foe in ip.n O'*T !T??"; .;nin uur ney

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L. oscms" sysen.T C g,g AtW3 SURVEILLANCE REQUIREMENTS / b L.\\ S I 3*l l .[. Each hydrogen monitor shall be demonstrated b M CPERA

ad are of a CHANNEL CHECEast least once perQh,BLE Fan a==by <=. the_

pe. r--_n n s331 MTr6sdEt TEST ~asr ,sess oner ter 3r usis) and at least once per 92' days @

%)EST am by perfd1 ming a CHANNEL CALIBRATION resTo obta calibrat pnpointso. . -

a. Ze volume perce it hydrog , and 44M

b. ne voiaame perce it hyd gen.

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McGUIRE - UNIT 1 3/4 6-17 Amendment No. 166 O

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3. INSTRUMENTATION 3 3 9 REMOTE SHUTDOWN (NSTRUMfMTATI

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k M .3 3 The remote shutdown hioi-ine,nsu - : : - A -aners shoiBn in gg. a shall be OPERABLEQi~th reapetitsTspl3reo enernp to the cante4 l APPLICABILITY: MODES 1, 2, and 3. g 4 gg;, egf.) g ,, ,,u ,

ACTION: UI ' U"" E b * '

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[th - c. ..A kw Taht/t t_o) restore the d nBoir--ir **

enasnafT11_ *'$ "

Asb., 6 OPERABL status'within days, or be n HOT SHUTDOWN within g 5 " E g 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . 30 t,,g A+S64 m (,63 Q l

At $ The provisions of Specification 3.0.4 are not applicable, i

i SURVEILLANCE REQUIREMENTS O . SR ').3.4.1 a,J sg,3,5 M,3 W V Each remote shutdown monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECKtand CHANNEL ,

CALIBRATION operations at the frequencies shown in Table 4. .

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1 McGUIRE - UNIT 1 3/4 3-47 Amendment No. 166 O

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133hACCIDENT MrX!TORINGLINSTRtME*iT6JJQg ,

6 1MfTING CONDITJdN FOR OPE [Ord

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Tabl eerMent ~tM instrumentation (ctiannels sh6ilibin shall be OPERABLE.

G3~D (fnTCh APPLICABILITY: MODES 1, 2, and 3. F 5ege,4 WA*a e g,43

's et.we) Cre.A ad'a.

It HOCT ACFod A

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a. With the k1 channel er of T)PERABLE aycfdent monitoring ins 3rGmentation ess than the RequJred Number of Channelyshown in

/N5tRT Table -10.frestore e Inoperaone pnannetts) to OPERA 8LE statu pf Asiod O wt days, or at least H0VSHUTDOWN within/thgp Ifpc C lith the er of OPERABLE acc ent monitoring inst channel ess than the Mini ntat o Channels OPEFABLE Table .3-10, restore the i perablechannel(?) uirements of ace ding to b.1 or b.2 b ow: OPERABLE status 4 ,,

.1 Instruments 1-1 within 48 hou ' #

lNS @ 19 at least HOT DOWN withiri the nexTIZ hours'. Y b.2 n 6 and 17: ~according echnical Specificati I.p#E oJ f @ The provisions of Specification 3.0.4 are not applicable.

O QHSERT Ac00N, 4

$ SURVEILLANCE REQUIREMENTS 5 ' 1 3 3- 4 .-

o# Each accident monitoring instrumentation channel shall be f# $3'D demonstrated OPERABLE by performance.of the. CHANNEL CHECliaand CHANNEL- -

CALIBRATION operations at the frequencies shown in Table .3-7.

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eknnel thd a c nornwHyenergzed, __

McGUIRE - UNIT 2 3/4 3-50 Amendment No. 148 i

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fg IAINMENT SYST k M 3 '3 ~3 l I 3/4.6.4 C

IBLE GAS CONTROL

/' HYDROGE ITORS I L NG CONDITION FOR OP TION e

/ i 3.6A1) Two independent containment hydrogen monitors shall be OPERABLE. '

? 3 >l APPLICAB',LIII: MODES 1 $ A. M ACTION:

With one hydrogen monitor inoperable 1 restore the inoperable monitor to JtCatf ERLE 10UrGstatus - within 30 daygrorpn at least mySTANDBY withiytne next}

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SR 3.3.3 M Each hydrogen monitor shall be demonstrated hA"V OPERABLE _by ti.e j af er N3W RATIONAtruanceit.>iof at a CHANiEL warst onerner2 CHEChat ame andleast once at least onceper per.97 qLhsigr.f.M0G days 45iiE C@ T !

(STAGGEDED TESTADESIbby pejrming a CHANNEL CALIBRATION (/0 sing nyarg mixtures to obt n calibrat n point' ou

a. Ze volume perce t hydr gt.a and

b. ine volume perci nt drogen.

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5,3 INSTRUMENTATION

% % 4~ ,.2.y 3.'!LQ REMOTE SHUTDOWNdNST/UMENTAI4tfN

( LIMITING CONDITION FOR OPERATION v

[5yshs Ldie*3 ,;

uo 3.3.

S Table _ remote shutdownGI6nitoriero instrumentation channeM_shMin

&- shall be OPERABLE @ttn re3souts aisprayco cuccm to tneAontro_1 LA,2o APPLICABILITY: MODES 1. 2, and 3.

ACTION: __

-_ M*TE 2

6[& ea.4A is ako.iel * -

Men erSh4 tyn.r l i,,,,,g t E (G44 e5 )

Adm A I. (#ait'h t [hunoer of urtMp5Lt .6

,te snufdown monitarisfa channelt less)

A s vi 8 r_a Maby Table'3.3-3r resterd the ca^rr>nia. channsrlID) to' re M)

OPERABLE status within days, or be71n HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . nSe 61 6 6e6 T o {g g )

Acrtods The provisions of Specification 3.0.4 are not applicable. O

NOfC I i

i SURVEILLANCE REQUIREMENTS O-E3'3'II

I (.JRT. Each remote si;utdown monitoring instrumentation channel shall be Se 3.3.4.3 dtili5nstrated OPERAE.LE by performance of the CHANNEL CHEC and CHANNEL CALIBRATION operations at the frequencies st wn in Table ..

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Discussicn cf Ch:ng:s S;ctica 3.3 - Instrumentation

() TECHNICAL CHANGES - LESS RESTRICTIVE L.30 CTS 3.3.3.5 Table 3.3-9liststheReadoutLocationfo/ remote shutdown instrumentation. The change moves this level of detail information to plant procedures. This type of information is not necessary in the Technical Specifications. ITS 3.3.4 retains the !

controls which require operability of remote shutdown systems, j therefore, there is no reduction in equipment requirements within the TS and no reduction in any safety analysis assumptions. This change is consistent with NUREG-1431.

L.31 Not used.

L.32 The CTS surveillance requirements for the 3.3.3.6 PAM, 3.6.4.1 Hydrogen Monitors, and 3.3.3 5 Remote Shutdown System Specifications were revised consistent with the STS to only require the channel check surveillance to be performed on normally energized instrumentation. It is not the intent of the CTS or the STS to require that equipment or channels be energized or control transferred in order to perform this survetIlance. However, the performance of a Channel Check on de-energized instrumentation does not provide meaningful information for determining the status (G) of an instrument channel and is not a productive use of plant resources. Therefore, the ITS Channel Check surveillance in 3.3.3 PAM and 3.3.4 Remote Shutdown System include an exception for the performance of this surveillance on de-energized equipment. The exception provided in the ITS eliminates the existing requirement to perform a Channel Check on de-energized equipment. Based on the fact that the performance of a Channel Check on de-energized equipment does not yield useful information for determining the Operability of a channel and therefore does not contribute to the safe operation of the plant, the elimination of the requirement to perform such surveillance is acceptable and allaws for the more productive use of plant resources.

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V McGuire Units 1 and 2 Page L - 10 Supplement 65l

.N3 Significant Hazards Ccnsidtrcticn Szctien 3.3 - Instrumentation.

LESS RESTRICTIVE CHANGE L.3i

.Not used.. !

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O. i l=McGuireUnits.Iand2 'Page 53 of 59 Supplement 65

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N3 Signifiernt H':znrds C:nsid;ratica l S;cti n 3.3 - Instrumentation (O) LESS RESTRICTIVE CHANGE L.32 The McGuire Nuclear Station is converting to the Improved Tbchnical Specifications (ITS) as outIined in NUREG-1431, " Standard Technicol Specifications, Westinghouse Plants." The proposed change involves making the current Technical Specifications (CTS) less restrictive.

Below is the description of this less restrictive change and the No Significant Hazards Consideration for conversion to NUREG-1431.

The CTS surveillance requirements for the 3.3.3.6 PAM, 3.6.4.1 Hydrogen Monitors, and 3.3.3 5 Remote Shutdown System Specifications were revised l consistent with the STS to only require the channel check surveillance to be performed on normally energized instrumentation. It issout the intent of the CTS or the STS to require that equipment or channels be energized or control transferred in order to perform this survetIlance. ;

However, the perfv:mance of a Channel Check on de-energized instrumentation does not provide meaningful information for determining the status of an instrument channel and is not a productive use of plant resources. Therefore, the ITS Channel Check surveillance in 3.3.3 PAM l and 3.3.4 Remote Shutdown System include en exception for the performance of this surveillance on de-ene'gized equipment. The

( ) exception provided in the ITS eliminates the existing requirement to L perform a Channel Check on de-energized equipment. Based on the fact that the performance of a Channel Check on de-energized equipment does ,

not yield useful information for determining the Operabtiity of a l channel and therefore dc?s not contribute to use safe operation of the l plant, the elimination of the requirement to perform such surveillance is acceptable and allows for the more productive use of plant resources.

In accordance with the criteria set forth in 10 CFR 50.92, the McGuire Nuclear Station has evaluated this proposed Technical Specifications change and detennined it does not represent a significant hazards consideration. The follo:ing is provided in support of this conclusion.

1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated? ,

I The proposed change removes the requirement to perform channel check survelIlonces on de-energized channels. This change will not affect the probability of an accident. The performance of l this channel check or the equipment involved are not initiators of analyzed events. The performance of a surveillance on de-energized l

{} channels does not provide on indication of equipment or instrument l

() operability. The consequences of an accident are not offected by J

McGuire Units 1 and 2 'Page 54 of 59 Supplement 65l l l l

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N3 Signifiednt Hucrds C nsider tien S2cticn 3.3 - Instrumentetien this change. ' The change continues to require the specific plant equipment to remain operable. Therefore, this change,will not involve a significant increase in the probability or consequence of an accident previously evaluated.

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

This change will not physically alter the plant (no new or !

different type of equipment wiil be installed). The changes in methods governt'ng normal plant operation are consistent with current safety analysis assumptions. Therefore, the change does not create the possibility of a new or different kind of accident' from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of

' safety?.

The margin of safety is. not affected by this change. The proposed

~

change does not alter the'TS requirement for this instrumentation

.to be operable and available to utilize in the mitigation of, or O recovery from, design basis events and continues to limit plant operatton when the instrumentation is not operable. The elimination of the-channel check surveillance on de-energized channels does not affect the safe operation of the plant. The applicable safety analysis assumptions continue to be maintained in a similar manner as before, therefore, the change does not involve a significant reduction in a margin of safety.

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McGuire & Catawba improved TS Review Comments Section 3.3 Instrumentation s

3.3.3-08 CTS 4.6.4.1 ITS SR 3.3.3.1 ITS SR 3.3.3.2 DOC L'11 The justification for the 31 day channel check frequency extension is that consistency is l' established with other PAM function TS frequencies. The justification for elim!nating operational tests is the passive nature of the instrument. Comment: Provide a technical basis l for proposed changes. Explain why the proposed ITS testing requirements are sufficient to

( ensure instrument operability.

1 DEC Response:

DOC L11 is revised to provide additional justification for the proposed changes.

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1( i 3 July 1,1998 mc3_cr_3.3a 2.

Discussi:n cf Ching::s S:cticn 3.3 - Instrument ticn

(. j TECHNICAL CHANGES - LESS RESTRICTIVE for monitoring radioactivity levels are readily availbble in nuclear plants and may be used as an adequate substitute for the installed radiation monitors. The TS Actions require the NRC to be informed by written report of the corrective Actions taken which would include a description of the alternate method used.

Therefore, the details of any alternate monitoring methods are

\ ject to further NRC review. Considering the adequacy and silable of alternate methods for monitoring radioactivity cnd

.e requirement to report the method used to the NRC, the use of alternate radiation monitoring methods is acceptable and may eliminate unnecessary plant shutdowns. This change is consistent with NUREG-1431.

L.10 The CTS 3.3.3.6 Action b Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months and CTS 3.7.4a Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is extended to 7 days for all channels, except hydrogen monitors. For hydrogen monitors, the completion time is extended from 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The extended completion time, retained in ITS 3.3.3, applies to single channel functions with the required channel inoperable and to two

(^)

U channel functions when both required channels of the applicable function are inoperable. Increasing the Completion Time for these instruments to 7 days (or 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for the hydrogen monitors) is acceptable because of the low probability of an event raauiring PAM instrument operation and the availability of alternate means to obtain the required information. This change is consistent with NUREG-1431.

L.11 CTS 4.6.4.1 for the Hydrogen Monitors requires a channel check once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , a monthly analog channel operational test, and a channel calibration 92 days on a staggered test basis. ITS 3.3.3 requires a channel check once per 31 days and a channel calibration once per 92 days. Elimination of the channel i

operational test and extension of the channel check frequency to l 31 days arch acceptable since the Hydrogen Monitors are passive devices; they do not initiate any automatic actuations, are not used for the routine operation of the plant, monitoring parameters important to reoctor safety, or in the mitigation of any design basis accidents. The Hydrogen Monitors and-are used only during post accident conditions in containment. The purpose of a channel operational test is de;igned to verify the required alarm, interlock, and trip functions of the instrumentation being tested.

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(~)'y L The Hydrogen Monitors do not have required alarm, interlock, or McGuire Units 1 and 2 Page L - 4 Supplement 65 l i

Discussicn of Ch:ng:s S:cticn 3.3 - Instrumentation h

(Q TECHNICAL CHANGES - LESS RESTRICTIVE tri;.o functions. In addition, the Hydrogen monitors hohe no routine monitoring function that would warrant frequent channel checks.

The Hydrogen Monitors are required in the TS to provide indication l of the post accident hydrogen concentration in the containment so '

that the Hydrogen Recombiners may be placed in operation when appropriate to limit hydrogen concentration. They are not required during or inanediately after a design basis event due to the relatively slow rate of hydrogen production after a design basis accident. Should the monitors fatI, manual sampling of the containment atmosphere and continous operation of the Hydrogen Recombiners would eliminate the need for the monitors. Therefore,

the channel operational test and frequent channel checks are not l Justified for the Hydrogen Monitors and do not significantly '

contribute to the continued operability of these monitors or to plant safety. The Hydrogen Monitor Operability continues to be l assured in the some manner as other indicating instrumentation required in the TS by the performance of a Channel Calibration \

every 18 months. In addition, tThe extension of the channei check i frequency from 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to 31 days is consistent with the frequencies established for all other post accident monitoring i instrumentation within the CTS. Therefore, this frequency for the channel check survelIlonce is adequate for the hydrogen monttors i as well. The elimination of the staggered testing is acceptable ,

since the monitors do not perform a mitigative function and this change avoids potential missed surveillance due to missing the- !

staggered intervals. Test data for these monitors indicates that the 92 day interval is sufficient to ensure operability. The hydrogen monitors are only required for monitoring and are not immediately needed after an event. This change is consistent with NUREG-1431.

L.12 CTS 3.3.3.5 Action a requires tha inoperable remote shutdown system instrument channeis to be restored to OPERABLE status within 7 days. ITS 3.3.4 increases the Completion Time from 7 days to 30 days. Thi's change is reasonable based on operating experience and the low probability of an event occurring that would require the control room to be evacuated. This change is consistent with NUREG-1431.

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lMcGuireUnits1and2 Page t - 5 Supplement 65

Na Significant Hazards Ccnsideraticn Szcticn 3.3 - Instrumentation LESS RESTRICTIVE CHANGE L.11 The McGuire Nuclear Station is converting to the Improved T$chnical Specifications (ITS) as outlined in NUREG-1431, " Standard Technical Specifications, westinghouse Plants." The proposed change involves making the current Technical Specifications (CTS) less restrictive.

Below is the description of this less restrictive change and the No Significant Hazards Consideration for conversion to NUREG-1431.

CTS 4.6.4.1 for the Hydrogen Monitors requires a ci:anael check once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , a monthly ana1og channel cperctional test, and 1 a channel calibration 92 days on a staggered test basis. ITS 3.3.3 requires a channel check once per 31 days and a channel calibration once per 92 days. Elimination of the channel operational test and extension of the channel check frequency to 31 days are4s acceptable since the Hydrogen Monitors are passive devices; they do not initiate any automatic actuations, are not used for the routine operation of the plant, monitoring parameters important to reactor safety, or in the mitigation of any design basis accidents. The Hydrogen Monitors ad-are used only during post accident conditions in containment. The purpose of a channel O operational test is de;igned to verify the required alarm, interlock, and trip functions of the instrumentation being tested.

The Hydrogen Monitors do not have required alarm, interlock, or trip functions. In addition, the Hydrogen monitors have no routine monitoring function that would warrant frequent channel checks.

The Hydrogen Monitors are required in the TS to provide indication of the post accident hydrogen concentration in the containment so that the Hydrogen Recombiners may be placed in operation when appropriate to limit hydrogen concentration. They are not required during or immediately after a design basis event due to the relatively slow rate of hydrogen production after a design basis accident. Should the monitors foil, manual sampling of the containment atmosphere and continous operation of the Hydrogen Recombiners would eliminate the need for the monitors. Therefore, the channel operational test and frequent channel checks are not justified for the Hydrogen Monitors and do not significantly contribute to the continued operability of these monitors or to plant safety. The Hydrogen Monitor Operability continves to be assured in the same manner as other indicating instrumentation required in the TS by the performance of a Chanrel Calibration every 18 months. In addition, tThe extension o' the channel check frequency from 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to 31 days is consistent with the O frequencies established for all other post accident monitoring lMcGuireUnits1and2 Page 25 of 59 Supplement 65

N) Significant Nrz2rds Ctnsidircticn Section 3.3 - Instrumentation instrumentation within the CTS. Therefore, this frequency for the channel check surveillance is adequate for the hydrogen monitors as well. The elimination of the staggered testing is* acceptable since the monitors do' rot perform a mitigative function and this change avoids potential missed surveillance due to missing the staggered intervals. Test data for these monitors indicates that the 92 day interval is sufficient to ensure operability. The hydrogen monitors are only required for monitoring and are not immediately needed after an ever.t. This change is consistent with

, NUREG-1431.

In accordance with the criteria set forth in 10 CFR 50.92, the McGuire Nuclear Station has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following is provided in support of this conclusion.

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

The proposed change revises the surveillance requirements for the hydrogen monitors. This change will not affect the probability of fJ w an accident. The hydrogen monitors are not initiators of any analyzed event. The role of the hydrogen monitors is to provide information to the operators during post accident conditions. The consequences of an accident is not affected by this change. The change will not alter assumptions relative to the mitigation of an accident or transient event. Therefore, this change will not involve a significant increase in the probability or consequence of an accident previously evaluated.

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

This change will not physically alter the plant (no new or different type of equipment will be installed). The changes in methods governing normal plant operation are consistent with current safety analysis ass ~umptions. Therefore, the change does not create.the possibility of a new or different kiad of aciident from any accident previously evaluated.

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McGuire Units 1 and 2 ' Page 26 of 59 Supplement 65 l

,h-Na Significant Haz:rds Ccnsid;raticn S:cticn 3.3 - Instrumentation ry h 3. Does this change involve a significant reduction in a margin of safety? ,

The hydrogen monitors provide a passive function (indication only) and the most common outcome of the performance of Surveillance is the demonstration that the acceptance criteria are met. Also, operating experience has shown that a 31 day channel check is sufficient to verify post accident monitoring instrumentation remains operable between calibration intervals. The safety

, analysis assumptions will still be maintained, therefore, the change does not involve a significant reduction in a margin of safety.

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b lMcGuireUnits1and2 'Page 27 of 59 Supplement 65

i. _____-_________-___--__A

l l McGuire & Catawba improved TS Review Comments Section 3.3 Instrumentation l

3.3.3-10 (McGuire only) ITS Conditions B DOC L 8

.lTS proposea to allow greater flexibility in plant operation by extending repair times to 30 days from 7 days for all two channel PAM functions. Currently, DOC L 8 states the justification is l based on operating experience and the remaining operable channels. Comment: Provide a detailed evaluation for staff review to allow it to determine the safety of the proposed change.

DEC Response:

I' i DOC L8 is revised to provide additional justification for the proposed change.

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\ mc3._cr_3.3a 4 July 1,1998

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Discussicn cf Ch:nges 5::cticn 3.3 - Instrumentation V TECHNICAL CHANGES - LESS RESTRICTIVE initiation of each indisidual pump is not a requirement of the safety analysis. Sufficient redundancy is provided by.the number

! of pumps in the system and the number of instrument channels available for automatic actuation retained in ITS 3.3.2. This change is consistent with NUREG-1431.

L.7 Not Used.

L.8 The CTS 3.3.3.6 Action a Completion Time allows 7 days for a l single inoperable channel of post accident monitoring (PAM) instrumentation for those instruments requiring two channels. ITS 3.3.3 allows 30 days for this condition. Increasing _the Completion Time to 30 days is acceptable based on operating experience and the remaining OPERABLE channels. These channels are passive and perform no actuation function. In general, the purpose and function of the McGuire PAM instrumentation is the some as the PAM instrumentation in the STS and in other Westinghouse designed plants. Therefore, when comparing the McGuire instrumentation to the STS, a detailed evaluation is not f required. The basis for the relaxation of the Completion Time for G' the McGuire PAM instruments is essentially the some as previously approved by the NRC for the STS. For two channel functions the redundant channel-is capable of monitoring the required PAM variable and remains ovatlable during the 30 day period.

Considering the availability of the redundant monitoring '

instrumentation to perform the required safety function and the low probability of an event occurring which would require the PAM instrumentation, the extension of the Completion Time to 30 days, I consistent with the STS, is acceptable. -This change is consistent l with NUREG-1431.

1 L.9 CTS 3.3.3.6 Action a and CTS 3.6.4.1 Action a require a unit '

shutdown when one required channel is inoperable and the Actions cannot be completed. ITS 3.3.3 allows continued operation in this condition for functions with two required channels provided a special report is written to the NRC detailing planned correct 1ve L actions. This change also applies to the containment radiation ,

l monitor function when the required channel is inoperable. This change is acceptable based on the remaining OPERABLE required

]

1 channel or, for the radiation monitor, this change is acceptable )

ps based on alternative methods of obtaining the required

.l information, such as taking samples. Temporary alternate methods i

L l'McGuireUnits1and2 Page L - 3 Supplement 65 l

N) Significant Hanrds C!nsid:ratien 5:cticn 3.3 - Instrumentation m.

(Q LESS RESTRICTIVE CHANGE L.8 The McGuire Nuclear Station is converting to the Improved Tdchnical Specifications (ITS) as outlined in NUREG-1431, " Standard Technical Specifications, Westinghouse Plants."- The proposed change involves making the current Technical Specifications (CTS) less restrictive.

Below is the description of this less restrictive change and the No 1 Significant Hazards Consideration for conversion to NUREG-1431 The CTS 3.3.3.6 Action a Completion Time allows 7 days for a single inoperable channel of post accident monitoring (PAM) instrumentation for those instruments requiring two channels. ITS 3.3.3 allows 30 days for this condition. Increasing the ]

Completion Time to 30 days is acceptable based on operating i experience and the remaining OPERABLE channels. These channels l are passive and perform no actuation function. In general, the l purpose and functton of the McGuire PAM instrumentation is the i same as the PAM instrumentation in the STS and in other i Westinghouse designed plants. Therefore, when comparing the !

McGuire instrumentation to the STS, a detailed evaluation is not i required. The basis for the relaxation of the Completion Time for l the McGuire PAM instruments is essentially the same as previously (Y approved by the NRC for the STS. For two channel functions the i

i redundant channel is capable of monitoring the required PAM ,

variable and remains avatlable during the 30 day period. l Considering the ovatIability of the redundant monitoring instrumentation to perform the required safety function and the l low probability of an event occurring which would require the PAM instrumentation, the extension of the Completion Time to 30 days, consistent with the STS, is acceptable. -This change is consistent with NUREG-1431.

In accordance with the criteria set forth in 10 CFR 50.92, the McGuire Nuclear Station has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards l consideration. The following is provided in support of this conclusion.

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

The proposed change increases the Completion Time when one required channel of PAM Instrumentation is inoperable from 7 days p to 30 days. This change will not affect the probability of an V accident. The PAM Instrumentation are not initiators of any lMcGuireUnits1and2 'Page 19 of 58 Supplement 65

No Significant Hazards Consideration Section 3.3 - Instrumentation analyzed events. The consequence of an accident is not affected by this change.

Post-accident indication of the requi, red parameter (s) is available from the remaining operable' channel.

The change will not alter assumptions relative to the mitigation of an accident or transient event. Therefore, this change will not involve a significant increase in the probability or consequence of an accident previously evaluated.

2.

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

This change will not physically alter the plant (no new or different type of equipment will be installed). The changes in methods governing normal plant operation are consistent with current safety analysis assumptions. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

} 3.

J Does this change involve a significant reduction in a margin of safety?

I The PAM instrumentation provides no automatic actuation functions.

The remaining required channel provides post accident indication and a timely corrective action period is specified for the inoperable channels. The safety analysis assumptions will still be maintained, therefore, the change does not involve a significant reduction in a margin of safety.

O McGuire Units 1 and 2 'Page 20 of 58 Supplement 65l

l N3 Significant H:zrrds C nsid:raticn Section 3.3 - Instrumentation 1

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10 / analyzed events. The consequence of an accident is not affected by this change. Post-accident indication of the requi, red parameter (s) is available from the remaining operable

channel.

The change will not alter assumptions relative to the mitigation of an accident or transient event. Therefore, this change will not involve a significant increase in the probability or consequence of an accident previously evaluated.

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2. Does the change create the possibility of a new or different kind l l , of accident from any accident previously evaluated? l This change will not physically alter the plant (no new or different type of equipment will be installed). The changes in methods governing normal plant operation are consistent with ;

current safety analysis assumptions. Therefore, the change does i not create the possibility of a new or different kind of accident from any accident previously evaluated. i

3. Does this change involve a significant reduction in a margin of safety? !

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! -(mbj The PAM instrumentation provides no automatic actuation functions.

The remaining required channel provides post accident indication l and a timely corrective action period is specified for the inoperable channels. The safety analysis assumptions will still {

be maintained, therefore, the change does not involve a significant reduction in a margin of safety.

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'v McGuire Units 1 and 2 'Page 20 of 58 Supplement 65 l L

I I McGuire & Catawba improved TS Review Comments

'_ Section 3.3 instrumentation D.3.3-11 ITS Condition D DOC L 9 l ITS proposes to allow greater flexibility in plant operations by allowing credit for diverse PAM L instrument functions. Comment: DOC L 9 does not explain why alternate monitoring methods l .or diverse channels provide a sufficient safety basis for the change to delete a shutdown requirement when inoperable channels are not repaired. The DOC also does not explain why l . diverse channels meet the intent of NUREG 1431 to include Regulatory Guide 1.97 Category l 1 and all Non-Category 1 Type A instruments in TS.

DEC Response:

. DOC L9 is revised to provide additional justification for the proposed change.

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/L 'mc3_cr_3.3a 5 July 1,1998 1

Discussi:n c,f Ch:ng:s S:cticn 3.3 - Instrumentation

{) TECHNICAL CHANGES - LESS RESTRICTIVE initiation of each individual pump is not a 'equirement of the safety analysis. Sufficient redundancy is provided by the number ;

of pumps in the system and the number of instrument channels l available for automatic actuation retained in ITS 3.3.2. This j change is consistent with NUREG-1431.

L.7 Not Used. 1 L.8 The CTS 3.3.3.6 Action a Completion Time allows 7 days for a ,

single inoperable channel of post accident monitoring (PAM) l instrumentation for those instruments requiring two channels. ITS 3.3.3 allows 30 days for this condition. Increasing the Completion Time to 30 days is acceptable based on operating ,

experience and the remaining OPERABLE channels. These channels ;

are passive and perform no actuation function. In general, the I purpose and function of the McGuire PAM instrumentation is the some as the PAM instrumentation in the STS and in other Westinghouse designed plants. Therefore, when comparing the McGuire instrumentation to the STS, a detailed evaluation is not 0

LJ required. The basis for the relaxation of the Completion Time for ,

the McGuire PAM instruments is essentially the some as previously approved by the NRC for the STS. For two channel functions the redundant channel is capable of monitoring the required PAM variable and remains Gvailable during the 30 day period.

Considering the availability of the redundant monitoring instrumentation tG perform the required safety function and the low probability of an event occurring which would require the PAM instrumentation, the extension of the Completion Time to 30 days, consistent with the STS, is acceptable. -This change is consistent with NUREG-1431.

L.9 CTS 3.3.3.6 Action a and CTS 3.6.4.1 Action a require a unit shutdown when one required channel is inoperable and the Actions cannot be completed. ITS 3.3.3 allows continued operation in this condition for functions with two required channels provided a special report is written to the NRC detailing planned corrective actions. This change also applies to the containment radiation monitor function when the required channel is inoperable. This change is acceptable based on the remaining OPERABLE required channel or, for the radiation monitor, this change is acceptable o based on alternative methods of obtaining the required

() l information, such as taking samples. Temporary altcrnate methods lMcGuireUnits1and2 Page L - 3 Supplement 65

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

Discussien cf Cha g;s Secticn 3.3 - Instrumentation 1

() TECHNICAL CHANGES - LESS RESTRICTIVE for monitoring radioactivity levels are readily avallo'ble in nuclear plants and may be used as on adequate substitute for the installed radiation monitors. The TS Actions require the NRC to be informed by written report of the corrective Actions taken which would include a description of the alternate method used.

Therefore, the detaiIs of any alternate monttoring methods are subject to further NRC review. Considering the adequacy and available of alternate methods for monitoring radioactivity and the requirement to report the method used to the NRC, the use of alternate radiation monitoring methods is acceptable and may eliminate unnecessary plant shutdowns. This change is consistent with NUREG-1431.

L.10 The CTS 3.3.3.6 Action b Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months and CTS 3.7.4a Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is extended to 7 days for all channels, except hydrogen monitors. For hydrogen monitors, the completion time is extended from 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The extended completion time, retained in ITS 3.3.3, applies to single channel functions with the required channel inoperable and to two

(_) channel functions when both required channels of the applicable v function are inoperable. Increasing the Completion Time for these instruments to 7 days (or 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for the hydrogen monitors) is acceptable because of the low probability of an event requiring PAM instrument operation and the availability of alternate means to obtain the required information. This change is consistent with NVREG-1431.

L.11 CTS 4.6.4.1 for the Hydrogen Monitors requires a channel check once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , a monthly analog channel operational test, and a channel calibration 92 days on a staggered test basis. ITS 3.3.3 requires a chaanel check once per 31 days and a channel calibration once per 92 days. Elimination of the channel operational test and extension of the channel check frequency to 31 days are4s acceptable since the Hydrogen Monitors are passive devices; they do not initiate any automatic actuations, are not used for the routine operation of the plant, monitoring parameters important to reactor safety, or in the mitigation of any design basis accidents. The Hydrogen Monitors ass-are used on1y during post accident conditions in containment. The purpose of a channel operational test is de:ig=d to verify the required alarm, p interlock, and trip functions of the instrumentation being tested.

Q The Hydrogen Monitors do not have required alarm, interlock, or McGuire Units 1 and 2 Page L - 4 Supplement 65l l

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l N2 Signific:nt H2zerds Ccnsid;retien l Srcticn 3.3 - Instrumentation LESS RESTRICTIVE CHANGE L.9 The McGuire Nuclear Station is converting to the Improved T$chnical

! Specifications (ITS) as outlined in NUREG-1431, " Standard Technical Specifications Westinghouse Plants." The proposed change involves making the current Technical Specifications (CTS) less restrictive.

Below is the description of this less restrictive change and the No Significant Hazards Consideration for conversion to NUREG-1431.

CTS 3.3.3.6 Action a and CTS 3.6.4.1 Action a require a unit

! shutdown when one required channel-is inoperable and the Actions cannot be completed. ITS 3.3.3 allows continued operation in this condition for functions with two required channels provided a special report is written to the NRC detailing planned corrective actions. This change also applies to the containment radiation monitor function when the required channel is inoperable. This change is acceptable based on the remaining OPERABLE required channel or, for the radiation monitor, this change is acceptable based on alternative methods of obtaining the required information, such as taking samples. Temporary alternate methods for monitoring radioactivity levels are readily available in

- O nuclear plants and may be used as on adequate substitute for the G installed radiation monitors. The TS Actions require the NRC to be informed by written report of the corrective Actions taken which would include a description of the alternate method used.

Therefore, the details of any alternate monitoring methods are '

subject to further NRC review. Considering the adequacy and available of alternate methods for monitoring radioactivity and the requirement to report the method used to the NRC, the use of alternate radiation monitoring methods is acceptable and may eliminate unnecessary plant shutdowns. This change is consistent with NUREG-1431.

In accordance with the criteria set forth in 10 CFR 50.92, the McGuire Nuclear Station has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following is provided in support of this conclusion.

1. Does the change involve a significant increase in the probability L

or consequence of an accident previously evaluated't The proposed change replaces the shutdown statement with a

( requirement to send a report to the NRC, outlining the preplanned V] alternate method of monitoring, the cause of the inoperability, lMcGuireUnits1and2 'Page 21 of 58 Supplement 65

N3 Significant H:zerds C nsid;ratica SIcticn 3.3 - Instrumentstien and the plans and schedule for restoring the instrumentation channels of the function to operable status. This change will not affect the probability of an accident since the PAM

instrumentation are not initiators of any analyzed event. Post-accident indication of the required parameter (s) 1s available from the remaining operable channel and from the preplanned alternate method of monitoring. The change will not alter assumptions relative to the mitigation of an accident or transient event.

Therefore, this change will not involve a significant increase in the probability or consequence of an accident previously evaluated.

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

This change will not physically alter the plant (no new or different type of equipment will be installed). The changes in methods governing normal plant operation are consistent with current safety analysis assumptions. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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% 3. Does this change involve a significant reduction in a margin of safety?

The proposed change replaces the shutdown statement with a requirement to send a report to the NRC, outlining the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to operable status. The margin of safety is not affected by this change because the remaining required channel or the preplanned alternate method of monitoring is available to provide the required indication. The PAM instrumentation provides no automatic actuation functions. The safety analysis assumptions will still be maintained. Therefore, the change does not involve a significant reduction in a margin of safety.

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U McGuire Units 1 and 2 'Page 22 of 58 Supplement 65 l L. s

l McGuire & Catawba improved TS Review Comments Section 3.3 instrumentation J

3.3.4, Remote Shutdown System l

! 3.3.4-02 CTS Table 3.3.9 DOC A 50 The CTS term "less than the Minimum Channels OPERABLE" is translated in ITS to "One or l more required Functions inoperable." This changes results in equivalent requirements because the CTS requires one channel to be operable for each TS function. Comment:

Revise A 50 to state the administrative nature of the CTS change.

DEC Response:

DOC A50 !s revised to provide additional justification for the proposed change.

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mc3_cr_3.3a 6 July 1,1998 f

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l Discussisn of Ching3s Section 3.3 - Instrumentation

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k ADMINISTRATIVE CHANGES and are more appropriately located in the post accident monitoring specification. The applicability and shutdown requirements of CTS 3.6.4.1 are rr. vised to be consistent with the more restrictive existing requirements in CTS 3.3.3.6. Technical changes are delineated in other discussions of changes. This change, retained in ITS 3.3.3, 1: consistent with NURE3-1431.

A.49 A Note is added to the CTS 3.3.3.5 Actions that allows separate condition entry for Functions in the Remote Shutdown System Table.

The Note, retained in ITS 3.3.4, provides enlicit instructions for proper application of the actions for Techi.tcal Specification compliance. In conjunction with ITS 1.3, " Completion Times," this Note provides direction consistent with the intent of the existing Actions for the Remote Shutdown Instrumentation. This change is !

administrative and is consistent with NUREG-1431. l A.50 CTS 3.3.3.5 Table 3.3-9 columns are combined into one column labeled, " Required Number of Functions." ITS 3.3.4 does not

! n retain the split table format and makes only one column header for Q- the existing " Minimum Channels OPERABLE" and " Total No. of channels" columns. This change specifies only the number of 1 required channels to be OPERABLE. The ITS lists the required channels as " functions" but specifies the same number as the CTS Minimum Channels Operable column. Additionally the ITS Actions are entered when one or more functions are inoperable in the same manner as the CTS Actions are entered with "less than the Minimum l Channels Operable". As such the reformat of the CTS Table does i not introduce a technical change in the effective number of required channels or functions listed or in the manner in which '

the Actions are entered. This change is administrative and is consistent with NUREG-1431. l A.51 CTS 3.3.2, " Engineered Safety Features Actuation System Instrumentation," contains the Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation. The ITS provides a separate Specification (ITS 3.3.5, " LOP DG Start Instrumentation") for this i function. This change also deletes the table format for the LOP l Instrumentation that currently exists. This change is administrative and is consistent with NUREG-1431.

O A.52 CTS 3.3.2 Table 3.3-3 APPLICABLE MODES for the LOP DG Start b Instrumentation are 1, 2, 3, and 4. ITS 3.3.5 requires the lMcGuireUnit1and2 Pa e A - 134-3 Supplement 65

t

, Discussien of Chrngis 52cticn 3.3 - Instrumentation ADMINISTRATIVE CHANGES instrumentation to be OPERABLE in MODES 1, 2, 3, and 4, and when associated DG is required to be OPERABLE by LC0 3.8.2, "AC Sources

-: Shutdown." CTS 4.8.1.2 requires certain surveillance from LC0 3.8.1.1 to be met, which includes' auto start capability.

Therefore, the addition'of this requirement is considered an administrative change. 'This change is consistent with NUREG-1431.

A.53 A Note is added to the CTS 3.3.2 Actions that allows separate condition entry for each LOP DG Start Function. The Note, retained in ITS 3.3.5, provides explicit' instructions for proper application of the actions for Technical Specification compliance.

In conjunction with ITS 1.3, " Completion Times," this Note provides direction consistent with the intent of the existing I actions for the LOP Instrumentation. This change is administrative and is consistent with NUREG-1431.

A.54 Not used.

A.55 Not Used.

D A.56 CTS 3.3.2, " Engineered Safety Features Actuation System Instrumentation," contains the Containment Purge and Exhaust-Isolation Instrumentation. The ITS provides a separate specification (ITS 3.3.6, " Containment Purge and Exhaust Isolation

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Instrumentation") for this function. This change is administrative and is consistent with NUREG-1431.

l- 'A.57 :A Note is added to the CTS 3.3.2 Actions that allows separate 4 - condition entry for each Containment Purge and Exhaust Isolation Instrumentation Function. ~This Note, retained in ITS 3.3.6, provides explicit instructions for proper application of the actions for Technical Specification compliance. In conjunction with ITS 1.3, " Completion Times," this Note provides direction consistent with the intent of the existing Actions for the ESFAS Instrumentation. This change is administrative and is consistent with NUREG-1431.

A.58-64 Not used.

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McGuire Unit 1:and 2- -Page A - 1444 Supplement 65l L

McGuire & Catawba improved TS Review Comments O Section 3.3 Instrumentation 3.3.4-03 DOC A 1 This justification is used to modify channel check requirements to apply only to those functions normally energized. As such, the proposed change may be administered, but A 1 designates format and wording changes. Comment: Provide a DOC for the substantive changes proposed for ITS channel check surveillance. Any changes that are not exclusively administrative will require an appropriate DOC.

- DEC Response:

DOC A.1 was replaced with DOC L.32. See response to question 3.3.3-07 (same issue) for marked-up pages and new DOC.

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g mc3_cr_3.3a 7 July 1,1998 9

McGuire & Catawba Impiceed TS Review Comments Section 3.3 Instrumentatbn

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3.3.4-04 CTS Table 4.3-6 1 DOC L 13 i

ITS Table 3.3.4-1, allows the use of either steam generator level or auxiliary feedwater flow l rates as remote shutdown functions to assess decay heat removal capability. Comment: l Provide an explanation of why the proposed change is safe.

DEC Response:

DOC L13 is revised to provide additional justification for the proposed change.

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( Discussion of Changes Section 3.3 - Instrumentation TECHNICAL CHANGES - LESS RESTRICTIVE L.13 CTS 3.3.3.5 Table 3.3-9 requires both auxiliary feedw$ter flow and steam generator level as separate indication of Decay Heat Removal via the SGs. ITS 3.3.4 allcws the use of either one or the other indicators rather than both. The purpose of these indications is to determine if decay heat removal is taking place via the SGs to ensure a safe shutdown. Therefore, adequate information to determine if decay heat removal is occurring via the SGs can be

, obtained by either of these indications. The AFW flow and SG level indicators are both RG 1.97 qualified instrumentation and will provide reliable information in a variety of plant conditions.

Each monitoring function is also required operable in the PAM LCO.

A single indication (either monitoring function) for one SG is adequate to determine that decay heat removal capability is available and each of these indications are required on all four SGs. As such, a single failure of one indicator out of four would not prevent the determination of adequate decay heat removal capability. Therefore, as these diverse indications provide on equivalent level of infonnation to determine the availability of the required decay heat removal capability, and are available on O all four SGs, the allowance provided by this change does not reduce the level of safety provided by the TS requirements and continues to assure on adequate level of information is maintained on the remote shutdown panels to ensure a safe remote shutdown.

Therefe, e, this change is acceptable. This change is consistent with NUREG-1431.

L.14 CTS SR 4.3.3.6 and Table 4.3-7 requires a CHANNEL CALIBRATION to be performed every refueling. ITS 3.3.3 contains a Note allowing the neutron detectors to be excluded from the CHANNEL CAL 1BRATION.

The CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. The test verifies that the channel responds to the measured parameter within the necessary range and accuracy.

The neutron detectors are excluded from the CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. This change is consistent with NUREG-1431.

L.15 Not used.

O McGuire Units 1 and 2 Page L - 6 Supplement 65l

l Discussien cf Ch:nges j S;cticn 3.3 - Instrumentation n

D TECHNICAL < CHANGES - LESS RESTRICTIVE l

L.16 CTS 3'.3.2 Table 3.3-3 Action 15a for LOP Instrumentation requires an entry into CTS 3.8.1.1 with more than one channel inoperable.

ITS 3.3.5 provides I hour to restore one channel to operable status. If the Required Actions for one channel inoperable, or for two or more channels inoperable are not met, the actions require entry into the applicable Condition (s) and Required Action (s) for the associated DG made inoperable by LOP

, Instrumentation. This action is reasonable to provide a limited time to correct minor inoperabilities prior to declaring the DG inoperable. If the channel cannot be repaired, the actions for an inoperable DG, LC0 3.8.1 or 3.8.2, provide adequate compensatory actions to assure plant safety. This change is consistent with NUREG-1431.

L.17 CTS 3.3.2 Action 17, for Containment Purge and Exhaust Isolation Instrumentation, allows operation to continue with one or more channels inoperable provided the containment purge valves are maintained closed. Since the CTS does not specify a completion time, it is assumed to be immediately. ITS 3.3.6 requires ITS LC0 f 3.6.3 to be entered immediately if one or more manual or automatic actuation trains ate inoperable. ITS LC0 3.6.3 allows 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to isolate the penetration with an inoperable containment isolation valve. The proposed Action is acceptable because the existing Actions of CTS 3.6.3 permit 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to close and isolate an inoperable containment isolation valve. The change provides consistency within the Technical Specifications for the requirements associated with these valves. This change is consistent with NUREG-1431.

, L.18-21 Not used.

L.22 CTS Table 3.3-3 Actions 15 and 15b allow operation to proceed with one inoperable channel (placed in trip) until the next performance

,. of the COT. ITS 3.3.2 Conditions D, J and P contain a note which l allows the channel to be placed in bypass for surveillance testing i on other channels. Performance of surveillance on other channels cannot be completed because the inoperable channel is in trip and cannot be taken out of trip without the note. Most j inoperabilities can be repaired prior to the performance of i surveillance on other channels. However, should a repair be l

delayed, a surveillance on an inoperable channel would be missed l McGuire Units. I and 2 Page L - 7 Supplement 65 l l

L_..___. . . _ _ _ . _.

j

Discussien sf Ch;ng:s S;ctien 3.3 - Instrumentation f%

C TECHNICAL CHANGES - LESS RESTRICTIVE and the operable channel would be declared inoperableI forcing a unit shutdown. This change is consistent with NUREG-1431.

L.23 Not used.

L.24. The RTS CTS Action 2c on Table 3.3-1, applicable to an inoperable power range channel, requires power to be reduced to less than or

, equal to 75% within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or QPTR must be monitored using R,e I movable incore detectors every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This CTS Action was revised consistent with the corresponding STS Action. The corresponding STS Action (D.1.2) provides consistency between the two options. The STS Actions require that either QPTR be verified or power be reduced to less than or equal to 75% within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The RTS safety function of the affected power range channel is satisfied by the Action requirement to place the channel in trip (CTS Action 2.a which has been retained as ITS Actions D.1.1 or D.2.1). The proposed ITS Actions continue to ensure the reactor trip safety function of the channel is accomplished. In j c addition, the slow change in QPTR over time allows an acceptable l level of safety to be maintained in the CTS and ITS by monitoring

(

V) QPTR every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months at a power level greater than 75%. Therefore, an acceptable level of safety is also provided by allowing a time I of no more than 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to reduce power to less than or equal to l 75% as an alternative Action to monitoring QPTR every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Therefore, the increase in the time allowed to reduce power to less than or equal to 75% from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is acceptable, j L.25 CTS Action 15 for Catawba and 15a for McGuire allow operation to !

proceed with one inoperable channel (placed in trip) until the l performance of the the next operational test. The CTS Action is j revised consistent with the STS to allow operation to continue .

indefinitely ance the channel is placed in trip. The trip logic associated with these channels provides an actuation when two out three channels on a bus are tripped. Therefore, the allowance to I continue operation with one inoperable channel in trip is ,

acceptable due to the fact that the channel in trip can be i i considered to have accomplished its safety function and the protection system for that bus is left in a conservative condition (only one channel out of the remaining two must trip in order to actuate the protection function). In addition, since two channels remain operable and only one is required to trip in order to actuate the protection function, the system is left in a condition s

l McGuire Units 1 and 2 Page L - 8 Supplement 65l

Discussicn of Ch:ngas S;cticn 3.3 - Instrumentation (3

() . TECHNICAL CHANGES - LESS RESTRICTIVE where no single failure would prevent the actuation of the protection function. As such the adoption of the STS Actions for '

these instrument channels continues to provide adequate assurance of safe plant operation.

L.26 Not used.

L.27 Not used.

L.28 CTS Action 26 for an inoperable channel of the Containment Pressure Control System on Table 3.3-3 is revised to simplify the required action. The CTS Action requires that the inoperable channel be placed in the start permissive mode and that the Actions for the applicable supported systems be entered within one hour. This CTS Action is revised to address one or more inoperable l channels and to simply declare the supported system inoperable immediately. The Containment Pressure Control System instrument I channels provide both a start permissive and a terminate function for the Containment Spray, Containment Air Return and Hydrogen

( ) Skinsner Systems. The terminate function provides protection ;

against inadvertent actuation and the resulting negative pressure l transient in the containment. If manually placed in one mode (as i required by the CTS Actions) the other safety function provided by the channel becomes unavailable. The proposed ITS Action does not require that one of the safety functions provided by the instrument channels be disabled. In addition, the proposed ITS Action provides adequate assurance that operation with any number of inoperable Containment Pressure Control instrument channels is limited consistent with the Comp 19 tion Times for the supported equipment. Therefore this chanrfe is acceptable and consistent with the ITS Actions for Can ua, L.23 CTS 3.3.3.6, Tables 3.3-10 and 'i.3-7 for Accident Monitoring Instrumentation, contains proceaural detail information which describes the plant unique identifier for some functions. This level of detail is not necessary within the Technical Specification and is deleted. ITS 3.3.3 retains the requirement for the equipment to be orerable, therefore, there is no reduction l in equipment requirements within the TS and no reduction in any l

safety analysis assumptions. This change is consistent with

[') NUREG-1431.

LJ lMcGuireUnits1and2 Page L - 9 Supplement 65

l N3 Signific:nt H:zcrds Crnsid:rati:n S:ctica 3.3 - Instrumentation p

i k LESS. RESTRICTIVE CHANGE L.13

)

TheMcGuireNuclearStationisconvertingtotheImprovedT$chnical

. Specifications (ITS) as outlined in NUREG-1431, " Standard Technical l Specifications Westinghouse Plants." The proposed change involves making the current Technical Specifications (CTS) less restrictive.

Below is the description of this less restrictive change and the No l Significant Hazards Consideration for conversion to NUREG-1431. !

i l

CTS 3.3.3.5 Table 3.3-9 requires both auxiliary feedwater flow and l

steam generator level as separate indication of Decay Heat Removal j via the SGs. ITS 3.3.4 allows the use of either one or the other '

indicators rather than both. The purpose of these indications is

to determine if decay heat removal is taking place via the SGs to ;

ensure a safe shutdown. Therefore, adequate information to l determine if decay heat removal is occurring via the SGs can be '

obtained by either of these indications. The AFW flow and SG level ind'cators are both RG 1.97 qualified instrumentation and will provide reliable information in a variety of plant conditions.

Each monitoring function is also required operable in the PAM LCO.

A single indication (either monitoring function) for one SG is adequate to determine that decay heat removal capability is 1 '

available and each of these indications are required on all four ,

l SGs. As such, a single failure of one indicator out of four would !

\ not prevent the determination of adequate decay heat removal l- capability. Therefore, as these diverse indications provide on equivalent level of information to determine the availability of the required decay heat removal capability, and are available on all four SGs, the allowance provided by this change does not l reduce the level of safety provided by the TS requirements and l l continues to assure an adequate level of information is maintained on the remote shutdown panels to ensure a safe remote shutdown.

Therefore, this change is acceptable. This change is consistent with NUREG-1431.

In accordance with the criteria set forth in 10 CFR 50.92, the McGuire Nuclear Station has evaluated this proposed Technical Specifications l change and determined it does not represent a significant hazards l consideration. The following is provided in support of this conclusion.

I k}

t' i

. i McGuire Units 1 and 2 'Page 30 of 59 Supplement 65 l l

No Significant Nazards Csnsid2ratien 5:cticn 3.3 - Instrumentation

1. 'Does' the change involve.a significant increase in the probability 1 or consequence of an accident previously evaluated?

j The proposed change will allow the use of either the auxiliary feedwater flow or the SG water level indications for the Decay Heat Removal vid the SGs Function. The change does not affect the probability of an accident.- The Remote Shutdown Instrumentation are not' assumed to be an' initiator of any analyzed event. The consequences of an accident are not affected by this change. The

, purpose of_these indicators is to determine if adequate decay heat removal is occurring. Adequate indication is provided by eithe'r of these instruments. This change will not alter assumptions relative to'the mitigation of an accident or transient. Therefore.

-the change will not involve a significant increase in the probability or consequence of an accident previously evaluated.

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

Thi_s change will not physically alter the plant (no new or different type of equipment will be installed). The changes in methods governing normal plant operation are consistent with current safety analysis assumptions. The proposed change will allow the use of either the auxiliary feedwater flow or SG water level instruments for the. Decay _ Heat Removal'via the SGs Function.

'Therefore, the. change does not create.the possibility of a new or .

different kind of. accident from~ any accident previously evaluated.

'3. Does this' change involve a significant reduction in a margin of safety?

!. The proposed change will allow the use of either the auxiliary feedwater flow or SG water level for the Decay Heat Removal via the SGs Function. The margin of safety is not affected by this change. Adequate decay heat removal indication will remain by allowing the use of either instrument. The safety analysis assumptions will still be maintained, therefore, the change does -

not involve a significant reduction in a margin of safety.

O~

-lMcGuireUnits1and2' Page 31 of 59 Supplement 66 c--_ -___ _ _ = - _ _ _ - - - . - - - - - - - - - _ - - _ - - - _ _ . . - _ _ _ - - _ - - - - - . - - - - _ _ _ - - - - - - - - - - _ . - - _ - - - _ . . - _ - - _ - _ - - _ - - - - - -

i McGuire at Catawba Improved TS Review Comments O

V Section 5.0, Administrative Controls i

l 5.3, Unit Staff Qualifications 5.3.1 Regulatory Guide (RG) 1.8, Revision 2, was released concurrent with the 1987 change to 10 CFR Part 55 and describes an acceptable means of meeting the rule. The Statement of Consideration for the Part 55 rule change states, "Those facility licenses that have made a commitment that is less than that required by the new rules must conform to the new rule automatically." The Statement of Consideration for the Part 55 rule change further states,

" Details regarding other training and qualification will not be required to be supplied on Form NRC-398, if these requirement are contained in an NRC-approved training program that uses a simulation facility acceptable to the NRC under 6 55.45(b). Subject to continued Commission endorsement of the industry's accreditation process under the Final Policy Statement on Training and Qualification of Nuclear Power Plant Personnel (50 FR 11147; March 20,1985), a facility licensee's training program would be approved by being accredited by the National Nuclear Accrediting Board."

Comment: When the NRC endorsed the industry's accreditation process as an acceptable alternative to p;oviding specific information on the license application, the eligibility guidance in the accredited training programs was considered to be equivalent to the NRC criteria in RG p 1.8, Revision 2. Your proposed requirement, on the other hand, references Rev. O of RG 1.8,

(~ dated September 1975. Please describe how your proposed requirement which is different from RG 1.8, Revision 2 meets the intent of the Statements of Consideration for 10 CFR Pari

55. Alternatively, you may consider revising your proposed requirement.

1 DEC Response:

The staff comment does not appear to be related to the actual TS content. CTS 6.3.1 and ITS 5.3.1 make reference to RG 1.8, Rev. O only with regards to the Radiation Protection Manager and makes no reference to operator licensing and training programs.10 CFR 55 is related to the requirements for licensed operators and does not relate to the Radiation Protection Manager position or other staff positions. Requirements for operator training and accreditation are not addressed by ITS 5.3.1 nor by the STS 5.3.1. CTS 6.4.1 does discuss the operator training and replacement programs and explicitly states that it complies with 10 CFR 55. However, this section is deleted from the iTS consistent with NUREG 1431 on the basis that the requirements of the regulation are directly implemented and enforceable. DEC does not believe that the proposed requirements for the Radiation Protection Manager and reference to RG 1.8, Rev. O for that position are in any way related to 10 CFR 55 and believes that the proposed ITS 5.3.1 for personnel qualifications is consistent with current licensing basis.

I (q/

mc4_cr_5.0 21 July 7,1998 l

I

1 s l

ENCLOSURE 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION O SECTION 3.5 l

l 9

i L. __ _ _ _ - - - . -

i McGuire & Catawba improved TS Review Comments (n] ITS Section 3.5, Emergency Core Cooling Systems )

1 1

3.5.1-02 Bases JFD 4 Bases discussion of the Applicable Safety Analyses for ITS 3.5.1, page B 3.5-3.

The Bases discussion for STS 3.5.1 quantifies the delay time used in the analysis to account for SI signal generation. The Bases discussion for corresponding ITS 3.5.1 has not adopted this material, but rather describes additional factors that contribute to the delay in ECCS flow.

Comment: No specific justification has been provided for omitting the quantification of the delay time. Revise the submittal to provide the appropriate delay time.

DEC Response:

The statement in the STS Bases is misleading and implies that the only delay time is for Si i signal generation. As noted, several factors are considered, each with a delay time, however, the STS only states the delay associated with Si signal generation and ignores those considerations with much longer delay times. The Bases were revised to indicate the various j considerations that constitute the delay time and acknowledge that the accumulator is the only source of flow during this time. These C 'sy times, e.g., Si signal delay, DG start, and pump starts, are identified and controlled by the response time testing requirements and are not necessary for inclusion within the accumulator Bases. JFD 7 was added to justify this

( change.

l REVISED RESPONSE:

I l The ITS Bases are revised to conform to the STS Bases and JFD 7 is deleted.

'(

k mc4_cr_3.5 2 June 29,1998 1

Accumulators B 3.5.1 BASES (v) l APPLICABLE this event, the accumulators discharge to the RCS as soon as

! SAFETY ANALYSES RCS pressure decreases to below accumulator pressure.

l (continued)

As a conservative estimate, no credit is taken for ECCS pump flow until an effective delay has elapsed. This delay accounts for the diesels starting, the valves opening, and the pumps being loaded and delivering full flow. The delay time is conservatively set with an additional 2 seconds to account for SI signal generation. During this time, the

, accumulators are analyzed as providing the sole source of l l emergency core cooling. No operator action is assumed l during the blowdown stage.of a large break LOCA. ,

l 1 The worst case small break LOCA analyses also assume a time l l delay before pumped flow reaches the core. For the larger ;

i range of small breaks, the rate of blowdown is such that the increase in fuel clad temperature is terminated solely by the accumulators, with pumped flow then providing continued cooling. As break size decreases, the accumulators, safety injection pumps, and centrifugal charging pumps all play a part in terminating the rise in clad temperature. As break l size continues to decrease, the role of the accumulators continues to decrease until they are not required and the l (em*)

! centrifugal charging pumps become solely responsibic for l terminating the temperature increase.

t This LC0 helps to ensure that the following acceptance criteria established for the ECCS by 10 CFR 50.46 (Ref. 3) ,

will be met following a LOCA: '

l

a. Maximum fuel element cladding temperature is s 2200*F; l

b. Maximum cladding oxidation is s 0.17 times the total cladding thickness before oxidation;

c. Maximum hydrogen generation from a zirconium water reaction is s 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding ;

cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react; and j

d. Core is maintained in a coolable geometry.

Since the accumulators discharge during the blowdown phase

of a LOCA, they do not contribute directly to the long term O (continued)

U l McGuire Unit 1 .B 3.5-3 Supplement 6 l l

l

Accumulators B 3.5.1 8ASES APPLICABLE The limiting large break LOCA is a double ended guillotine SAFETY ANALYSES . break at the discharge of the reactor coolant pump. During (continued) . this event, the accumulators discharge to the RCS as soon as

.RCS pressure. decreases to below accumulator pressure.

As a conservative estimate, no credit is taken for ECCS pump flow until an effective delay has elapsed. .This delay accounts for the diesels starting, the valves opening, and the pumps being loaded and delivering' full flow. The delay time is conservatively set with an additional 2 seconds to account for SI signal generation. During this time, the accumulators are analyzed as providing the sole source of emergency core cooling. No operator action is assumed ~

during the blowdown stage of a large break LOCA.

The worst case 'small break LOCA analyses also assume a. time delay before pumped flow reaches the core. For the larger range of small breaks, the rate of blowdown is such that the increase in fuel clad temperature is terminated solely by the accumulators, with pumped flow then providing continued cooling. As break size decreases, the accumulators, safety injection pumps, and centrifugal charging pumps all play a part in terminating. the rise in clad temperature. As break O size continues to 6ecrease, the role of the accumulators continues to decrease until they are not required and the centrifugal. charging pumps become solely responsible for

. terminating the temperature increase.

This LC0 helps to ensure that the following acceptance criteria established for the ECCS by 10 CFR 50.46 (Ref. 3) will be met following a LOCA:

a. Maximum fuel element cladding temperature is s 2200*F;

b. Maximum cladding oxidation is s 0.17 times the total cladding' thickness before oxidation;

c. Maximum hydrogen generation from a zirconium water reaction is s 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react; and i

d. Core is maintained in a coolable geometry.

[ (continued)

McGuire Unit 2 .B 3.5-3 Supplement 6 l

i Justification fcr Deviaticns l S:cticn 3.5 - Emerg:ncy Cera Cooling Systems (ECCS) t )

\m / BASES i

NOTE: The first five justifications for these changes fr$m NUREG-1431 l:

were generically used throughout the individual Bases section markups. Not all generic justifications are used in each section. 1 j

1

1. The brackets have been removed and the proper plant specific l information or value has been provided.

1

2. Editorial change for clarity or for consistency with the Improved Technical Specifications (ITS) Writer's Guide.

3. The requirement / statement has been deleted since it is not applicable to this facility. The following requirements have been renumbered, where applicable, to reflect this deletion.

4. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.

(q)

5. Bases have been modified to reflect changes made to the f admical specifications.

6. Westinghouse Nuclear Safety Advisory Letter (NSAL)97-003 has recommended a review of the ITS Ba es for accumulator isolation valves (NUEEG 3.1.5) based on shutdown LOCA considerations. These recommendations of the NSAL relate to the consideration of operating bypasses and compliance with IEEE 279-1971. The automatic open feature of these valves is not considered an operating bypass as described in the UFSAR since the valves are manually opened and power is removed.

The Bcses of ITS 3.5.1 has been revised consistent with the UFSAR and current licensing basis.

7. Not usedA; noted in STS 3.5.1 Base: Applieable Safety Analysis, :cveral delay facter; are considered in ECCS fic;;, however, the STS-Bases only tate; the delay a;;ccicted .ith SI signal generation and ignere; thc;c eens&deration: ith much longer delay times. The ITS 3.5.1 Ba;c; were revi;cd to delete the ;pecific SI-signal delay ahich i; nisicading-and indicat-c the varica; consideration; that constitute the total delay.

Speci fic delay time;, e.g., SI signal delay, DG-start, and pump tc%

are identified and controlled by re pon;c time tc;tivequ;rement; and

( ) are not necc;;ary for inclu;icn ithin t-he-accumulator Ba:cs.

/

I lMcGuireUnits1and2 114 Supplement 64

Accumulators B 3.5.1

~

BASES J

APPLICABLE SAFETY ANALYSES As a conservative estimate, no credit is taken for ECCS pump flow until an effective delay has elapsed. This delay .

(4

( inued) accounts for the diesels startingYand the pumps beino loade[dW. valves opaQ

.and deliverinc full f ow. The delay time is conservatively seMwitan aeditionf 2 secrWRinto account for SI signal ^~

bra 5Y gas pro <v;1 orcthe iding Durino this time, sole source the accumulators of emergency are No core cooling. analvm1 i F mr -y }

operator action is assumed during '.he blowdown stage of a Ive. i l large break LDCA. * * -

~

The worst case small break LOCA analyses also assume a time i delay before flow reaches the core. For the larger 1 range of smal breaks, the rate of blowdown is such that the l increase in fuel clad temperature is terminated solely by !

the accumulators, with pumped flow then providing continued cooling. As break size decreases, the accumulators /and b) g 9i hs + y S I l

centrifugal charging ptmps([ED play a part in terminating N y the rise in clad temperature. As DreaK size continues to 1 decrease, the role of the accumulators continues to decrease until they are not required and the c ntrifugal charging ptmps become solely responsible for terminating the <

temperature increase. '

This LCO helps to ensure that the following acceptance criteria established for the ECCS by 10 CFR 50.46 (Ref. 3) will be met following a LOCA:

a. Maximum fuel element cladding temperature is s 2200*F:

b. Maximum cladding oxidation is s 0.17 times the total k cladding thickness before oxidation: ,

c. Mcximum hydrogen generation from a zirconium water reaction is s 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react; and ;

d. Core is maintained in a coolable try.

of a LOCA. they do not contribute to b long fterm H aw,nebW

. Since the accumulators discharge dur coolingde

N aa,-A A blowd D4 requirements of 10 CFR 50.46. v_r_ -

heks b MM We.

c re c.m sidvhl For both the large and small break IDCA analyses, a nominal o44 rett oJ, Ah

. contained accumulator water volume is used. The contained cl.4. *o.g hi,,, g I

u e w y seurez (continued) b "'Mu b P A eA .' % A M g ,

B 3.5 3 Rev 1. 04/07/95 McGee O

v I

f a

.j b !

g

. McGuire & Catawba improved TS Review Comments h- ITS Section 3.5, Emergency Core Cooling Systems u

3.5.1-03 DOC L.1 CTS 3.5.1.e CTS 4.5.1.1.2

' CTS 3.5.1.e requires that a water level and pressure channel be Operable for each 1 accumulator. CTS 4.5.1.1.2 provides the Surveillance Requirements for these instrument

" channels. These requirements have not been adopted in corresponding ITS 3.5.1 in conformance wnh the STS and are being deleted from the CTS. ' Comment: These requirements are important enough to be maintained in a licensee controlled document.

Revise the' submittal to move these requirements to a location with appropriate change control, for example, the SLC. i DEC Response:

' DEC disagrees that additional control of these_ instruments is required outside of the technical

' specifications. The requirement of concem b the quantity of water and nitrogen pressure within each accumulator. There are a number of ways of determining compliance. with this requirement. . This may include using the installwl instrumentation, or it may include the use of {

' portable test equipment. Existing controls already ensure that equipment used to determine surveillance limits are appropriately calibrated. It would be an uneccessary administrative

burden to add an additional layer of control for these specific instruments when they are not i

- required nor may not even be used to determine compliance with the TS limit. There are i flarge numbers of instrumentation used throughout the plant to determine that various limits "r ; are w ti n hi acceptance criteria.' These instruments are adequately controlled without being

'w ' included within the SLC.'

REVISED RESPONSE: l l

The staff indicated during the comment resolution meeting that the proposed response was not acceptable and that the instruments must be relocated to the SLC. DEC disagrees. The 1 proposed instruments do not perform a safety function and are provided for indication. These l indicators are also not required as category,1 or type A post accident monitoring variables. l Therefore, these instruments do not meet any of the criteria established in 10 CFR 50.36. .

j.

The staff position that the instruments be maintained in a licensee controlled document is l already met by their inclusion within plant calibration procedures. The addition of these .

l

. Instruments to the SLC does not establish any additional control because no actions would be l required ivhen the instruments were inoperable. The addition of these instruments would l create an additional administrative burden on the plant staff by adding an unecessary l additional control document.' l 9

mc4;cr 3.5: 3 June 29,1998 '

?

IN k'

uy- I 'Y

McGuits & Catawba improved TS Rwiew Comments ITS Section 3.5, Emergency Core Cooling Systems i

3.5.3, ECCS - Shutdown 3.5.3-01 ' A.11 McGuire only CTS 4.5.3.1 ITS SR 3.5.3.1 Note

' The Note to STS SR 3.5.3.1 states, "An RHR train may be considered Operable during alignment and operation for deMy heat removal, if capable of being manually realigned to the ECCS mode of operation." This Note has been adopted in corresponding ITS SR 3.5.3.1.

DOC A.11 has categorized this proposed change to the CTS as administrative. Comment:

This proposed change is less restrictive. Revise the submittal to provide the appropriate justification for this proposed change.

DEC Response:

DEC diagrees that this change is less restrictive. The NRC has clearly identified this change as an administrative change in the Safety Evaluation for Vogtle Electric Generating Plant related to Amendment 96 (unit 1) and Amendment 74 (unit 2) for the conversion to improved technical specifications. The justification provided in DOC A.11 is consistent with the NRC conclusions and justifications provided by VEGP.

) REVISED RESPONSE: l l

The staff indicated during the comment resolution meeting that the proposed response was l not acceptable and that the change must be classified as either a less restrictive or more l restrictive change. DEC disagrees. The note was added to the ITS by the industry during the l proof and review period (July 1992) for revision 0 of NUREG-1431 and was identified as the l movement of bases information to a note to " aid in interpretation." The existing SER for l VEGP also concurs that the change is an administrative clarification. The existing CTS SR l 4.5.3.1 only requires the " applicable" requirements of CTS SR 4.5.2 be performed. This note l Is an administrative clarification that is intended to clarify that certain MODE 1 - 3 alignment l requirements for the RHR system are not applicable in MODE 4. The RHR system is l routinely operated in MODE 4 for decay heat removal and requires a change in alignment l' from the ECCS configuration in MODES 1 - 3. This does not represent a change in operating l

- methods and is consistent with all other plants which use the RHR system fcr both decay heat l removal and ECCS functions. DEC believes that clarifying note in ITS is consistent with the l CTS allowances that only the " applicable" requirements be performed and does not represent l any departure from the current licensing basis or system operation as described in the l UFSAR. Additionally, for McGuire, the proposed note is already contained in CTS 4.5.2.a l footnote, therefore, the ITS note is an administrative reformatting of existing requirements. l l

\ mc4_cr_3.5 7 June 29,1998

I

,- McGuire & Catawba improved TS Review Comments ITS Section 3.5, Emergency Core Cooling Systems i

3.5.3-02 DOC A.11 McGuire only JFD 8 CTS 4.5.3.1 ITS SR 3.5.3.1 Note Bases discussion for ITS SR 3.5.3.1, page B 3.5-23 The Note to STS SR 3.5.3.1 states,"An RHR train may be considered Operable during alignment and operation for decay heat removal, if capable of being manually realigned to the ECCS mode of operation." The Note to corresponding ITS SR 3.5.3.1 includes an additional provision for pressure isolation valve testing. JFD 8 states that realignment of the RHR system is necessary to establish the prerequisite conditions for the required testing of SR 3.4.14.1 in Mode 4. However, Note 1 to ITS SR 3.4.14.1 states, "Not required to be performed in Modes 3 and 4." Comment: The justification for the difference provided by JFD -

8 appears to be in error. Additionally, the proposed difference does not appear to be plant specific -it seems generic. Revise the subm.'tal to conform to the STS.

DEC Response: l l ,

1 The allowance in ITS SR 3.4.14.1 to operate in MODES 3 and 4 does not preclude testing in y these MODES. Actual testing is started on some valves in MODE 4 and continued for others

,j in MODE 3, therefore, the allowance is necessary in both MODES. The proposed note is consistent with the existing licensing basis in the footnote to CTS 4.5.2.a as described in DOC l A.11 and the SR 4.0.4 exception to CTS 3.4.6.2.

l REVISED RESPONSE:

l l JFD 8 only addresses the changes to the STS associated with required testing for the RHR l valves which is done in MODE 4. ITS SR 3.4.14.1 applies to all RCS PlVs and testing for l l PlV.s in o'her systems is also done in MODE 3. Therefore, there is no error or inconsistency l l

l in JFD 8 since it addresses testing for a single system rather than for all systems containing l l PlVs. l l- !

l 4 l I i

l

/~'N

' (.-) mc4_cr_3.5 6 June 29,1998

__________m_____. . _ _ . _ _ _ _ . . . . _ . _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ . - ___. . _ _ _ _ _ _ _ _ _ _ _ _. ._____________.____________________J

McGuire & Catawba hoproved TS Review Comments ITS Section 3.5, Emeripecy Core Cooling Systems E

3.5.3-03 CTS 3/4.5.3, Action b Action b of CTS 3/4.5.3 addresses the Condition of no ECCS Operable because of the inoperability of either the RHR heat exchanger or RHR pump. The CTS does not specify a time limit for implementing the Action. A revision has been proposed for the CTS that would i specify a Completion Time of immediately which conforms to the STS. Comment: No specific justification has been peovided for this proposed change. Revise the submittal to provide the justification for this change.

DEC Reeponse:

~ The CTS markup identifies this change as an A.1 change. The existing action to restore or provide attemate heat removal means without a specific time limit is interpreted as a requirement to initiate action without delay, i.e., immediately. There is no intent to change any requirements, therefore, the change is considered purely one of format and presentation covered by A.1 REVISED RESPONSE: l I

DOC M5 has been added to indicate that a specific time limit for the action is a more l restrictive change.

'l h/

N m

6 '? /

l 0 - mc4_cr_3.5 9 June 29,1998 O

e s h__ ___________m.______---_ _ - - - - _ . - - - - - - - - - - - - - - - - - - > - - - - - - - - - - - - - -- - - - - - - - - - - - - - - - - - ' - - - ' - - - - - ' ^ - ~ ~ - - - - ^ - - - - - - - - ' - ^ - - - -

SP E ba.3.d EMERGENCY CORE COOLING SYSTEMS (

^

t7215.3 ECCSMTEMS -4 </350Mhd2wh (TIiiETm.rwnulIIUR PUR OPliRATIQB i (_cn 3.5.3 Ef_*EITiiv~it;) one ECCS C --o s er o t the followin} shall be

.L OPERABLE:

~-

a. One OP LE centrifugal e rging -

b. On PERA8LE RHR heat e hanger, I

c. e OPERA 8LE RHR p and -

i . . An OPERABLE flow th capable of ta ng suction from the fueling I water storage ta upon being manu ly realigned and tr sferring

( suction to the ontainment sump d ing the recirculate phase of rea tion.

( APPLICA81LITY: MODE 4.

ACTION:

OPERABLE because 4-Q

!of the Inoperability of Ag@ @W0): no EhEC the centrifugal chargingt-~-- - -- n=T, ate r. - snm i j di

%.um =arar unmarr 2, restore er i net on1 ECCS subsystem to l gM., c OPERABLE status with'n I hour or be in COLD within the next .

AMa A g) 5CC OPERkB'L' ecause of no r a

@rtheRHRaest- - = = = -arem ----

subsystem to OPERABL. statusfor moin6 tore bar fdst M ECCS augnor sociarpt syst O c.

(L.7 sess snan asv r w use of alte w.

te heat INinoval methdds In the e~ vent tn cCS 'is actuatedAnd injects w r into the Coolant Syst a!pecialReporVshallbeprep and submit d to

.the Comissi pursuant to S fication 6.9. within 90 day

(,($ describin he circumstances f the actuati and the tota accumu-lated ac atton cycles to e. The cui value of the sage fac-or for each affected Saf y Injection n zie shall be ovided i t s tal Report whenever its value xce g o.70.

.)0 f A maximum of one centrifugal chargi pump and one Safety In.iection pump shall be capable of injecting into e RCS whenever the temperature of fWeD one or more of the RCS cold legs is less than or equal to 300*F. Two TTS 1.%.vt charg ng pumps may be operable and operating for s 15 minutes to allow

.y swarp ng charging pumps. Additional requirements are provided by Spect ication 3.4.9.3 McGUIRE - UNIT 1 3/4 5-7 Amendment No. 166 O

3 as

$6 cam M,3 EMERGENCY C_0RE COOLING SYSTDtS((6CCSh 32.5.3 ECt3 funwTnes . C < 350J) M

[LIMITigCONDfTffW40R OPFRM Lco 3-5.3 CAs fatnip6mlone ECC ys ear compriseVof the feb6winq} sha11 he OPERABLt:

[ One OPERA 8 entrifugal cha ng pump,f

b. One OPE LE RHR heat exc ger, A.7.

l c. One ERABLE RHR pump, nd

d. , OPERABLE flow pa capable of t ing suction f l the refuelin water storage tank pon being man 11y realigned a transferrin suction to the c. .ainment sum ring the rectre ation phase f operation.

I APPLICABILITY: MODE 4.

ACTION:

g pm3Q ,

%8@ With no ECC ys OPERABLE because1of the inanerability of either the centrifugal charging (RALup Df the f 36 path fyte D s*Yn.1wra warer.r.orage r=9 restore di me menu.c3 subsystem to 'I

, Ag c OPERABLE status w thin 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in TDOWN within th* ext h .

y,,, "a m aaksy m%

I no ECCS OPERABLE because of the ino i of Ach ' A @ d$ the RHR 4taganchanger.Jtt_Bilg pund restore ECCS subsystem to OPERABLE statustor maintain Tae Reactor Q6o ant sys

~

CJ Jtss than 357F Dy usef alternate / eat removal / methods.

c. In the event he ECCS is act ed and injects ter into the actor

' Coolant Sys , a Special Re rt shall be pre ared and subai ed to the Commis on pursuant to ification 6.9 2 within 90 da i

{,[) describin the circumstance of the actuati and the tota accumu-l lated ac ation cycles to ate. The curre t value of the sage fac-

' tor for ach affected sa ty Injection n zie shall be pr vided in !

hhisS ial Report whe ver its value ceeds 0.70.

1 l

I f

A maximum shall be capableof one of centrifugal charginbpump injecting into t RCS wheneverand one Saf'ety injection the temperature of pump g one or more of the RCS cold legs is less than or equal to 300*F. Two

' D*N.TL N charg ng pumps may be operable and operating for s 15 minutes to allow swapp

' LB g- chargi'ng pumps. Additional requirements are provided by peci cation 3.4.9.3

. McGUIRE

UNIT 2 3/4 5-7 Amendment No. 148 i

O -

Pe 3

9 l

l L ..

Discussicn of Ching s 5:ction 3.5 - Emergency Core Cooling System (ECCS)

TECHNICAL CHANGES - MORE RESTRICTIVE valve full open and specifies an explicit allowance and time limit for completing the surveillance. These additional requirements are more restrictive, however, they are consistent with the assumptions in the safety analysis and with NUREG-1431.

M.5 CTS 3/4.5.3 action b requires action to restore a subsystem when no ECCS train is operable because of the inoperability of either the RHR heat exchanger or RHR pump, but does not specify a time limit for implementing the Action. ITS 3.5.3 requires action to be immediately initiated. The proposed change is more restrictive since the CTS does not specify a time limit for starting the restoration, however, the change is consistent with the current implementation and interpretation for this action. This change is consistent with NUREG-1431.

,0 V

i O i McGuire Unit 1 and 2 Phge M - 23 Supplement 64 l l

L_________.- . _ _ - . _ - --

McGuire & Catawba improved TS Review Comments O ITS Section 3.5, Emergency Core Cooling Systems

?

3.5.4, Refue% Water Storage Tank 3 3.5.4-01 DOC A.15 JFD 6 j CTS 4.5.5.b I ITS SR 3.5.4.1

. CTS 4.5.5.b requires verifying that the solution temperature of the RWST is within limits once

= per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the outside air temperature is either less than 70*F or greater than 100*F.

Corresponding ITS SR 3.5.4.1 has a Frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months regardless of the outside air i temperature.' This proposed change has been categorized as administrative. Comment: This proposed change is more restrictive. Revise the submittal to provide the appropriate

, justification for this proposed chance.

DEC Reepones:

The submittialis revised to delete DOC A15 and reflect the proposed change as more restrictive change M3.

REVISED RESPONSE: l The staff indicated during the comment resolution meeting that the proposed DOC M'3 needed l additional clarification to show that the change was more restrictive. DOC M3 has been l revised to provide this additional clarification. l i

' mc4_cr_3.5 jJ June 29,1998 9

Discussign cf Ch:nges Section 3.5 - Emergency Core Cooling System (ECCS)

TECHNICA'L CHANGES - MORE RESTRICTIVE M.1 CTS 4.5.2.a permits valves to be realigned from requihed positions '

when placing the RHR system in service and for PIV testing for CTS 4.4.6.2.2. ITS 3.5.2 limits this time for realignment for PIV testing to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and allows these realignments only in MODE 3.

A two hour limit is placed on any configuration that isolates, at a given time, both trains of Safety Injection. The present requirement places no time limit on the testing. Limiting the time to two hours should provide sufficient time to perform the requirements and the time is consistent with NUREG-1431.

M.2 CTS 4.5.1.1.c requires that the power be removed from the accumulator isolation valve when RCS pressure is greater than 2000 psig. ITS SR 3.5.1.5' requires that the power be removed when RCS pressure is greater than 1000 psig. Westinghouse Nuclear Safety Advicory Letter (NSAL)97-003 has recommended changes to the Bases based on shutdown LOCA considerations. These recommendations I relate to the consideration of operating bypasses and compliance with IEEE 279-1971. The automatic unblock feature of these valves i p is not considered an operating bypass as described in the UFSAR

() since the valves are manually opened and power is removed. The Bases of ITS 3.5.1 has been revised consistent with the UFSAR.

The accumulator isolation valves are manually opened when RCS pressure is greater than 1000 psig. The requirement to remove I power to the valves at 1000 psig is more restrictive, but is consistent with existing practice and with the operability assumptions described in the subject NSAL for shutdown LOCA.

M.3 The alle=nce in CTS surveillance 4.5.5.b te-only requiresverify RWST temperature be verified when the ambient temperature is outside the limit. This allowance has been deleted in ITS SR 3.5.4.1. The same effort is required to routinely verify outside ambient temperature as is necessary to verify RWST temperature.

Therefore, the plant has elected to delete this allowance and perform ITS SR 3.5.4.1 every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This change is considered more restrictive since it involves the removal of an existing

' allowance.

l M.4 CTS-Surveillance 4.4.6.2.1.c requires the measurement of

[ controlled leakage at an RCS pressure of 2235 1 20 psig and

! provides an exception to the requirements of SR 4.0.4 in MODES 3 I and 4. ITS SR 3.5.5.1 requires that the seal injection flow lMcGuireUnit1and2 Page M - la Supplement-64 lL

.m McGuire & Catawba improved TS Review Comments

( ). ITS Section 3.5, Emergency Core Cooling Systems 3.5,' Additional Changes l 3.5-04 CTS 3.5.2 action b, CTS 3.5.3 action c l LA3 l

l In review comments for several sections, the NRC has identified that the deletion of l requirements which are redundant to regulation should be classified as less restrictive

. l changes rather than removal of detail changes. Therefore, DOC LA3 is deleted and replaced l by DOC L13 to be consistent with other changes made to other sections and to eliminate the l relocation of a requirement to a procedure.

O L

mc4_cr_3.5 20 June 29,1998 9

Sici AMa., 3.r,1.

1 EMERGENCY CORE COOLING SYSTEMS [(GCMO MX). 5. 2 ECCS EDBSYSTEMS - Ig, a 350 k]

CIN!J ins umulTION/DR OPERAT109 /

f., l 4. C8 3.5.2 Two (Ia"*aandeWEmergency coreAcolier tv< ten YCCSf, shegg be OPERABLE mi .. .. ....y>6==

mi ... v a 3

a. One 0P LE centrifugal charging p ,

b. One OP BLE Safety Injection pump

c. On PERABLE RHR heat exchanger

d. e OPERABLE RNR pump, and

e. An OPERABLE flow path ca e of taking suctio from the refuelin water storage tank en a fety Injection sign and automatica11 transferring suction t the containment suuy during the A.l d g recirculation phase o operation. _

APPLICABILITY: MODES 1, 2, and 3. _

Et BAS l

ACTION:

IaA4 fb. eqva4 wt.INF h a s100%

d d +i e ore *^

f Arki A @ (With on inoperable t r[t . in perable L( 4 to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be ir at least HOT STANDBY Ath.B within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWI within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

O . In the event th CCS is actues u .... acm weter into the act Coolant System a Special Report sha be prepared and subai ed to the Consissio pursuant to Speciff ion 6.9.2 within 90 d s describing e circumstances of actuation and the tot accumu-

[* O lated act ton cycles to date. The current value of t usage fac-tor for ch affected Safety jection nozzle shall b rovidedinj

. is S ial Report wheneve its value exceeds 0.70.

4 McGUIRE - UNIT 1 3/4 S-3 Amendment No. 166 O

,' , .: i oO Y

. sjekka.n3 g EMERGENCY CORE C00llNG SYSTEMS ( M _

I

( 3@)5.3 ECCSM-& </350MSIkd2 MTimytunUIIIUR tUR OP5 RAT 109

! [ g 3.5.3 FA 'si><fimps;) one ECCS r. unwoserot the foll_ggni)shall be

. k OPERABLE:

a. One OP LE centrifugal e rging -

b. On PERABLE RHR heat e hanger, I

' c. e OPERABLE RHR p and

. . An OPERABLE flow th capable of ta ng suction from the fueling water storage ta upon being ma'nu ly realigned and tr sferring t the ontainment sump d ing the recirculate phase of J

APPLICABILITY: MODE 4.

W11 : no EC OPERABLE because te flity of S)ilih the centrifugal chargingCgLi - - timatr r.

tne 6Pa*1im warar wrrowtanp, restore at i net on3 ECCS subsystem to Ac g., c OPERABLE status with' n I hour or be in C0 DOWN within the next l '

,o {.,,d.'.kh d'.M AM- A @ no EC OPEPABLEi*causeofth no C i tv nN

the RHR641rFttiniWIRcarEiu n"= i tore t M ECCS a As tor i.oolaag syst O c.

a su em to OPEPABLE status r6F = sin 6 less cnon n u r y use of alte In the event tn CCS is actuate te heat val meth6ds nd injects wa r into the R o Coolant S st a Special Repo hall be prep and submit d to

,the Commi si pursuant to Spe fication 6.9. within 90 day ,

[,,,[3 describir he circumstances f the actuati and the tota accumu-

. lated ac ation cycles to e. The curre value of the sage fac-or for each affected Saf y Injection n zie shall be ovided i t s Special Report whenever its value xceeds 0.70-i

,lo # A maximum of one centrifugal charging pump and one Safety Injection pump h

shall be capable of injecting into the RCS whenever the temperature of yweve TD one or more of the RCS cold legs is less than or equal to 300'F. Two rp 3,g n, charg ng pumps may be operable and operating for :s;15 minutes to allow

.y swapp ng charging pumps. Additional requirements are provided by Spect ication 3.4.9.3 McGUIRE - UNIT 1 3/4 5-7 Amendment No. 166 {

o '

p , .o

Spchwhen 'S S. 2.

EMERGENCY CORE COOLING SYSTEMS ((f cc5))

3fA.S.2 ECCS dUBSYSTE M - Tm / 50*F)

~'

KINIT18di CONDITIOI(FOR OPERATI M '

Lco 3.5.2 Twofindenendedt EmergencyA ore Cooling 5yst p (ECCS shall be OPERABLE wttn en suosystem % .we vu

a. One OP RA8LE centrifugal harging pump, l

b. One ERABLE Safety Inj tion pusy,

c. OPERABLE RHR heat xchanger,

d. OPERABLE RHR , and

e. 'An OPERABLE flow th capable of takin suction from the fueling

, water storage ta on a Safety Inject on signal and out tically transferring suc fon to the contai t sump during th A. ,, recirculation se of operation.

APPLICABILITY: MODES 1, 2, and 3. 1 g g gcc3 g :- ow

. O.., 34.u 6 s @ OPfAAst I eccs s . w e 1

, g- 4A @ With one ECCS Kubsyf noperabl restore the inoperable dune m ent to LE status ws 7Z hou or be in at least HOT STAND 8Y ,

g within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

)

fb. In the event the E is act'uated and i ects water into the acto 7 Coolant System, pecial Report shall prepared and subai ed to O

the Commission rsuant to Specificat 6.9.2 within 90 describing the circumstances of the tuation and the tot accumu-lated actuati cycles to date. Th current value of th usage fac-s i J

tor for eac affected Safety Injec on nozzle shall be vided in ,

this Spect Report whenever its alue exceeds 0.70.

i McGUIRE - UNIT 2 3/4 5-3 Amendment No. 148 ,

J O ..

l

, l hjt I +4 S C_____________________________-.-__ __ ]

pc6m .M.3 EMERGENCY CORE COOLING SYSTEMSh6CC.Sh -

N 5.3 ECCS f(iNKM"rFMS - C < 350Jh ILIMITI I CONDITinN 40R OPFR M LCe 3:5.3 (As fotnip6m3one EC vsteur compriseFof the fou6winalsha11 be OPERABLt E One OPERABL entrifugal char ng pump,#

b. One RHR heat exc r.

A

c. One ERABLE RHR pump,

d. OPERABLE flow pa capable of t ing suction f the refuelt ter storage tank being na 11y realigned a transferri suction to the c ineent sump operation. ring the rectre ation phase j l

APPLICABILITY: MODE 4.

M8 [=bsys4h

% 8 @ Withelther no EC the OPERABLE because.of tho (-rability of .

r1f I che ing qtump se the f' si path fa n %

Nr = . restore a wie anew.t.ca subsystem to gc status w thin I hour or be in within the next-hou . *2 gag inh A di.n A @ no ECCS.____ _$_ _

OPERABLE era;iility of (12 the RHR dieaF: -' 7 n_rbecause m pung, of the inop# Tl"i34RDECCS restore i i

(,,,;ubsystem to OPERABLE statusi9r samma ice neactor (4e ant sys j

)tss than 357F Dy usep alternate / eat removal / methods.

O Cc.

N In the event Coolant sys , a special ECCS is act ed and injects ter into the rt shall be p red and subai ed actor'I to the Commis on pursuant to ification 6.9 2 within 90 da

[, $ describi the circumstance of the actuati and the tota accumu-lated ac ation cycles to te. The curre t value of the sage fac-och affected Sa ty Injection zie shall be p vided in (torfor this S ial Report who ver its value ceeds 0.70.

I ChDI f A maximum of one centrifugal charging pump and one Saf'ety Injection pump shall be capable of injecting into the RCs whenever the temperature of j one or more of the RCS cold legs is less than or equal to 300*F. Two l'

.tt charg ng pumps.may be operable and operating for s 15 minutes to allow swapp ng charging pumps. Additional requirements are provided by

'Oc Speci ication 3.4.9.3 f . McGUIRE - UNIT 2 3/4 5-7 Amendment No. 148 l

l O .

a , <. >

)

I

Discussien cf Chingis Section 3.5 - Emergency Core Cooling System (ECCS)

TECHNICAL CHANGES - LESS RESTRICTIVE concentration or temperature will be slow and if the fimits are exceeded it will probably not be by a significant amount. This change.is-also acceptable based on the low probability of an event occurring:that'could not be properly mitigated. This change is consistent with NUREG-1431.

~

! L.10 CTS 3.5.1.1 action b for an accumulator inoperable due to a closed isolation valve requires the valve to be opened immediately or a shutdown be initiated. CTS 3.5.1.1 Action a for an accumulator inoperable due to other reasons provides'I hour to correct the L . inoperability prior, to requiring a unit shutdown. ITS 3.5.1

~

l Action B combines these requirements and provides 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to correct inoperabiliti'es for reasons other than boron concentration. The additional I hour completion time for unisolating an accumulator provides a reasonable time to attempt restoration activities and.has a negligible impact on safety.

This change.is consistent with NUREG-1431.

L.11 'Not used.

O L.12 ~ CTS 3.4.6.2.e requires that RCS controlled leakage be within l . limits during operation in MOCES 1-4.- ITS 3.5.5 only requires that the limits be maintained in MODES 1-3. The deletion of the Mode 4 requirement for Controlled Leakage (ITS seal injection h flow) is acceptable based on seal injection flow limits being less critical to the overall ECCS flow assumed in the applicable safety.

analyses due to the. lower initial RCS pressure and decay heat removal requirements in this Mode. The change is also acceptable because the specification only applies with RCS pressure between 2215 and -2255 psig. This range'of pressure can only be obtained in Modes 1, 2, and 3. .This change is consistent.with NUREG-1431.

L.13 The CTS 3.5.2 action b and CTS 3.5.3 action c requirements to provide a special report for ECCS actuation is redundant to regulation and is deleted. 10 CFR 50.73(a)(2)(iv) requires a report in.the event of an ECCS actuation within 30 days. The CFRs

. ore directly enforceable and are not necessary for duplication within the TS. The CFR provides sufficient regulatory control for this requirement. .There is no impact on safety since the requirements of the CFRs are directly implemented. This change is f consistent with NUREG-1431.

'McGuire Unit 1 and'2 Page L'- 33 Supplement 65/20/97l

Discussion of Changes Section 3.5 - Emergency Core Cooling System (ECCS)

TECHNICAL CHANGES - REMOVAL OF DETAILS LA.1 Not Used.

! LA.2 The detail of what individual subsystems constitutes an OPERABLE ECCS subsystem, contained in CTS Lf0 3.5.2 and 3.5.3, is being moved to the Bases of ITS 3.5.2 and 3.5.3. This level of detail is not required within the TS. The requirements contained in ITS 3.5.2 and 3.5.3 provide adequate controls for system operability.

Changes to the ITS Bases are subject to the controls described in ITS Chapter 5, " Administrative Controls." This change is consistent with NUREG-1431.

LA.3 Not usedThe CTS 3.5.2 action b and CTS 3.S.3 action c.requ w t+

tc provide ; pecial ' report for ECCS cctuation i; =ved to plant precedurc; for regulatcry reporting. 10 CFR 50.73(c)(2)(iv) require; a report in the event of an ECCS actuatien .ithin 30 W- The exi; ting regulatory rr,uirc=nt; previde sufficient control Over this report ithcut providing unnece;;;ry duplicatica

.ithin the TS. Thi; change i; consi; tent . tith NUREC l'31.

LA.4 The details of the method used to assure that the ECCS piping is full of water, contained in CTS Surveillance Requirement 4.5.2.b.1, are being moved to the Bases of ITS 3.5.2.3. The procedural details of how the surveillance is perfermed is not necessary for inclusion within the TS. The Bases are controlled in accordance with ITS Chapter 5.0, " Administrative Controls", and require a 10 CFR 50.59 evaluation. These controls ensui that changes to the Bases are appropriately reviewed. This change is consistent with NUREG-1431.

LA.5 Details of pump pressure limits for inservice testing specified in CTS 4.5.2.f are being moved to plant controlled documents for Insu'vice Testing Program requirements applicable to the ECCS pumps. The Inservice Testing program is controlled by 10 CFR 50.55a providing adequate assurance that any proposed change to the relocated requirements will be adequately reviewed prior to implementation. This change is consistent with NUREG-1431.

LA.6 The CTS 3.5.3 action b requires alternate heat removal methods if an RHR subsystem cannot be restored. This requirement has been moved to the Bases of ITS 3.5.3 and the action has been modified to ihinediately initiate action to restore. This change is appropriate since the time to complete the restoration is unknown l McGuire Unit 1 and 2 Page LA - 123 Supplement 65/E0/97

N3 Signific:nt H:zards Consid:rati:ns Section 3.5 - Emergency Core Cooling System 7,

LJ\ LESS RESTRICTIVE CHANGE L.13 The McGuire Nuclear Station is converting to the Improved Technical Specifications (ITS) as outlined in NUREG-1431, " Standard Technical Specifications, Westinghouse Plants." The proposed change involves making the current Technical Specifications (CTS) less restrictive.

Below is the description of this less restrictive change and the No Significant Hazards Consideration for conversion to NUREG-1431.

The CTS 3.5.2 action b and CTS 3.5.3 action c requirements to provide a special report for ECCS actuation is redundant to regulation and is deleted. 10 CFR 50.73(a)(2)(iv) requires a report in the event of an ECCS actuation within 30 days. The CFRs are directly enforceable and are not necessary for duplication within tie TS. The CFR provides sufficient regulatory control for this requirement. There is no impact on safety since the requirements of the CFRs are directly implemented. This change is consistent with NUREG-1431.

In accordance with the criteria set forth in 10 CFR 50.92, the McGuire Nuclear Station has evaluated this proposed Technical Specifications i

change and determined if does not represent a significant hazards consideration. The foilowing is provided in support of this conclusion.

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

The proposed changes delete requirements from Technical Specificottons which are already specified in the CFR. The location of regulatory controls is not an initiator of any analyzed event, therefore, the proposed change does not offect the probability of any analyzed accident. The CFR is directly implemented and enforceable. Therefore, the requirements associated with the CFR will continue to be met. The safety analysis assumptions associated with analyzed events are not offected by the source location of regulatory requirements, therefore, the conseq~ences u of analyzed event 5 are not changed.

i

/

\_

l

McGuire Units 1 and 2 Page 2629 of 3029 Supplement 64 l

1.

No Signiikc:t Hazards C nsidersticns Section 3.5 - Emergency Core Cooling System l

,Q'% ) 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? ,

The ' change wiil not physicolly alter the plant (no new or different type of equipment will be installed). The changes in methods governing normal plant operation are consistent with current safety anu!ysis assumptions. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident prev ously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The proposed changes delete requirements from Technical Specifications which 'are already specified in the CFR. The changes do not reduce the margin of safety since the CFR is directly implemented and enforceable and the location of regulatory requirements has no imp'3ct on any safety analysis assumptions.

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l'McGuireUnits1and2 Page 2729 of 3029 Supplement 64

McGuire & Catawba improved TS Review Comments ITS Section 3.5, Emergency Core Cooling Systems 3.5, Additional Changes ,

3.5-05 ITS SR 3.5.1.5, page 3.5-3 l

ITS Bases SR 3.5.1.5, page B 3.5-8 l STS SR 3.5.1.5 and associated bases

' l f I

CTS SR 4.5.1.1.1.c requires verification that power is removed to each accumulator isolation l

valve operator when the RCS presscr- is > 1000 psig. STS 3.5.1.5 and associated Bases l use " pressurizer" pressure is > 1000 psig. Changing " pressurizer" to "RCS" is corrected to be l consistent with the current system design. l l .

l i

O

Accumulators 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.5 Verify power is removed from each 31 days accumulator isolation valve operator when RCS'is > 1000 psig. l l-1 0

McGuire Unit 1 , 3.5-3 Supplement 6 l E__---_----_-----_-------_---------------------------------- ----- - -------a

Accumulators 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.5 Verify power is removed from each 31 days accumulator isolation valve operator when RCS pressure is > 1000 psig. l O

O

'McGuire Unit 2 , 3.5-3 Supplement 6 l u-__------____----------_- - _ - - . - - - - - - - - - - - - - - - - -)

Accumulators B 3.5.1 BASES SURVEILLANCE SR 3.5.1.4 (continued) I REQUIREMENTS necessary to verify boron concentration if the added water inventory is from the refueling water storage tank (RWST),

because the water contained in the RWST is within the accumulator boron concentration requirements. This is consistent with the recommendation of NUREG-1366 (Ref. 6).

SR 3.5.1.5 Verification every 31 days that power is removed from each accumulator isolation valve operator using the power l disconnect switches in the correct position when the RCS pressure is > 1000 psig ensures that an active failure could not result in the undetected closure of an accumulator motor operated isolation valve. If this were to occur, only two accumulators would be available for injection given a single failure coincident with a LOCA. Since power is removed under administrative control, the 31 day Frequency will provide adequate assurance that power is removed.

p' This SR allows power to be supplied to the motor' operated

( l isolation valves when RCS pressure is s 1000 psig, thus allowing operational flexibility by avoiding unnecessary delays to manipulate the breakers during plant startups or shutdowns. Even with power supplied to the valves, inadvertent closure is prevented by the RCS pressure interlock associated with the valves.

Should closure of a valve occur in spite of the interlock, the SI signal provided to the valves would open a closed valve in the event of a LOCA.

i (continued)

LJ l McGuire Unit l' . B 3.5-8 Supplement 6

Accumulators B 3.5.1 BASES SURVEILLANCE SR 3.5.1.4 I (continued)

REQUIREMENTS necessary to verify boron concentration if the added water inventory is from the refueling water storage tank (RWST),

because the water containea in the RWST is within the accumulator boron concentrate requirements. This is consistent with the reconsnendation of NUREG-1366 (Ref. 6).

SR 3.5.1.5 Verification every 31 days that power is removed from each j accumulator isolation valve operator using the power j l disconnect switches in the correct position when the RCS i pressure is > 1000 psig ensuret that an active failure could I not result in the undetected closure of an accumulator motor operated isolation valve. If this were to occur, only two accumulators would be available for injection given a single failure coincident with a LOCA. . Since power is removed under administrative control, the 31 day Frequency will provide adequate assurance that power is removed.

q This SR allows power to be supplied to the motor operated tg l isolation valves when RCS pressure is s 1000 psig, thus allowing operational flexibility by avoiding unnecessary delays to manipulate the breakers during plant startups or shutdowns. Even with power supplied to the valves, inadvertent closure is prevented by the RCS pressure interlock associated with the valves.

Should closure of a valve occur in. spite of the interlock, the SI signal provided to the valves would open a closed valve in the event of a LOCA.

i i

(continued) !

V(N

'l McGuire Unit 2 , 8 3.5-8 Supplement 6 I

l

Accumulators 3.5.1 SURVEILI.MCE REQUIREMENTS (continued)

SURVEILLANCE FRE R Y SR 3.5.1.5 Verify power is removed from each 31 days e=h+= isolation valve rator when i

b .

l 9

l l

e h

9 I

3.5 3 Rev 1. 04/07/95

h1COtwN

.e

Accumu1ators B 3.5.1

( y SURVEILLANCE SR 3.5.1.4 ,

REQUIRDENTS *

(continued) The boron concentration should be verified to be within required limits for each accumulator every 31 days sirv.e the static design of the accumulators limits the ways in which the cor m-n ation can be changed. The 31 day frequency is ,

adequate to identify changes that could occur from mechar: isms such as stratification or inleakage. Samp11po thir affected i accumulator within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a 124 volume incriiE will ' A Y identify whether inleakage has caused a reduction tn 24 con '

concentration to below the required limit. It is not necessary to verify boron concentration.if the ar'ded water inventory is from the refueling water storage tank (Rl6T).

because the water contained in the RWST is within the t accumulator boron cerehation requirements. this is I consistent with the recommendation of IGEG 1363 (Ref. g ]

SR 3.5.1.5 Verificati ev 000

^

31 da that power is removed free each Q .

h .2 accumulator sol pressure is ion psig ensuresvafve that operatorgenen an'actwo MWQ(T m E.M tg8% g""I could 5D I

i out. result iin tie undetected closure of an accumu1V,or motor) operated isolation valve. If this were to occur. only two ( 8"A(" M k -

i

accumulators would be available for injection given a single n f failure coincident with a LOCA. Since is removed under b

( administrative control, the 31 day will provide I adequate assurance that power is .

\ This SR allows power to be supplied to the ated isolation valves when pressurizer pressure is Stpsio.

thus allowing operational flexibility by avoi ng unnecessary @

delays to sanipulate the breakers during plant startups or shutdowns. Even with power supplied to the valves.

inadvertent closure is preve associated with the valves. nted by the RCS pressure interlock

\

Should closure of a valve occur in spite of the interlock, the -

SI signal provided to the valves would open a closed valve 1n the event of a LOCA. ,

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(continued) 8 3.5 8 Rev 1. 04/07/95 m cm, e'

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

McGuire & Catawba improved TS Review Comments ITS Section *J.5, Emergency Core Cooling Systems 3.5,' Addidonal Changes j l 3.5-06 ITS Bases Background, page B 3.5-r l STS Bases Background, insert page B 3.5-2 I

I-

- l . ITS Bases Background discussion of the isolation valves between the accumulators and the

~ l - RCS states that the valves must be " locked open" during unit operation. To be consistent l with further discussion of these valves in the STS Bases 3.5.1 LCO section, this discussion is l changed to reflect the valves as "open and power remover,".

O i.

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- d' . mc4_cr_3.5 22 June 29,1998 l I

Accumulators B 3.5.1 BASES BACKGROUND This interlock also prevents inadvertent closur$ of the (continued) valves during normal operation prior to an accident. The valves will automatically open, however, as a result of an L SI signal. The isolation valves between the accumulators and the Reactor Coolant System are required to be open and power removed during unit operation.. In that the subject valves are normally open and do not serve as-an active device during a LOCA, the requirements of the Institute of Electrical and Electronic Engineers (IEEE). Standard 279-1971 (Ref.1)-is not applicable in this situation. Therefore, the subject valve control circuit is not designed to this standard.

The accumulator size, water volume, and nitrogen.e 1 l pressure are selected so that three of the four at-nulators-are sufficient to partially cover the core before significant clad melting or zirconium water reaction can occur following a LOCA. The need to ensure that three accumulators are adequate for this function is consisteht with the LOCA assumption that the entire contents of one accumulator will be lost via the RCS pipe break during the blowdown phase of the LOCA.

O APPLICABLE The accumulators are assumed OPERABLE in both the large and SAFETY ANALYSES small break LOCA analyses at full power (Ref. 2). These are the Design Basis Accidents (DBAs) that establish the acceptance limits for the accumulators. Reference to the analyses for these DBAs is used to assess changes in the i accumulators as they relate to the' acceptance limits.

In performing the LOCA calculations, conservative assumptions are made concerning the availability of ECCS flow. In'the early stages of a LOCA, with or without a loss of offsite power, the accumulators provide the sole source i of makeup water to the RCS. The assumption of loss of offsite power is required by regulations and conservatively imposes a delay wherein the ECCS pumps cannot deliver flow until the emergency diesel generators start, come to rated speed, and go through their timed loading sequence. In cold leg break scenarios, the entire contents of one accumulator are assumed to be lost through the break.

The. limiting'large break LOCA is a double ended guillotine

- break at the discharge of the reactor coolant pump. - During

( (continued) l McGuire Unit.1 .B 3.5-2 Supplement 6

?

w-.- --___-_D_. --.i--._-___ - - - - - - _ - - . . . - - .b

Accumulators B 3.5.1 BASES

' BACKGROUND. This interlock also prevents inadvertent closure of the ,

(continued)' valves during normal operation prior to an accident. - The 1 valves will automatically open, however, as. a result of an SI signal. The isolation valves between the accumulators and the Reactor Coolant System are required to be open and power removed during unit operation. In that the subject valves are nonnally open and do not serve as an active device.during a LOCA, the requirements of the Institute of Electrical and Electronic Engineers (IEEE) Standard 279-1971 (Ref.1) is not applicable in this situation. Therefore, the subject valve control circuit is not designed to this standard.

The accumulator size, water volume, and nitrogen cover pressure are selected so that three of the four accumulators are sufficient to partially cover the core before significant clad melting or. zirconium water reaction can occur following a LOCA. The need to ensure that three accumulators are adequate for this function is consistent with the LOCA assumption that the entire contents of one accumulator will be lost via the RCS pipe break during the blowdown phase of the LOCA.

-I APPLICABLE The accumulators are assumed OPERABLE in both the large and SAFETY ANALYSES small break LOCA analyses at full These are.

the Design Basis Accidents (DBAs) that power (Ref.

establish the2).

acceptance limits for the accumulators. Reference to the analyses for these DBAs is used to assess changes in the accumulators.as they relate to the acceptance. limits.

In performing the LOCA calculations, conservative assumptions are made concerning the availability of ECCS flow. In the early stages of a LOCA, with or without a loss of offsite power, the accumulators provide the sole source of makeup water to the RCS. The assumption of loss of offsite power is required by regulations and conservatively imposes a delay wherein the ECCS pumps cannot deliver flow until the emergency diesel generators start, come to rated speed, and go through their timed loading sequence. In cold i

leg break scenarios, the entire contents of one accumulator are assumed to be lost through the break.

(continued) l McGuire Unit 2 .8 3.5-2 Supplement 6 a

1 INSERT O .The' isolation valves between the accumulators and-the Reactor Coolant System are required to be open and power removed during unit operation.

that the subject valves are.normally open and do not serve as an active Inl

)

device during a LOCA, the requirements of the Institute of Electrical and Electronic Engineers (IEEE) Standard 279-1971 (Ref. 1) is not applicable in this situation. Therefore,-the subject valve control circuit is not designed to this standard.

O IMe$mve.

INSERT Page B 3.5-2 9

O ENCLOSURE 3

/ RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION C] SECTION 3.6 I I

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McGuire & Catawba Improved TS Iteview Comments ITS Section 3.6, Containment Systems 3.6.1-2 DOC A.3 DOC LA.1 I

-. JFD 6 '

JFD Bases 6 CTS 4.6.1.1.c .

CTS 3.6.1.2.a. b, and c CTS 4.6.1.2.d CTS 4.6.1.2.d. 3)

CTS 4.6.1.2.e, h, and I

. ' CTS 4.6.1.6 STS SR 3.6.1.1 ITS SR 3.6.1.1 ITS B3.6.1 Bases

CTS 4.6.1.1.c, 3.6.1.2, 3.6.1.2 ACTIONS,4.6.1.2 and 4.6.1.2c, d, e, f, g, h and I specify various leak rate testing requirements and criteria for containment.- CTS 4.6.1.6 specifies visual examinations to be performed on the containment vessel. STS SR 3.6.1.1 requires the visual examination and leakage rate testing be performed in accordance with 10 CFR 50 Appendix J as modified by approved exemptions. ITS SR 3.6.1.1 modifies STS SR 3.6.1.1 to conform to TSTF 52. The STS is based on 10 CFR 50 Appendix J Option A while the ITS is based on.10 CFR 50 Appendix J Option B. Changes to the STS with regards to Option A versus Option B are ccvered by a letter from Mr. Christopher 1. Grimes to Mr. David J.-

G~ ' Modeen, NEl dated 11/2/95 and TSTF 52 as modified by staff comments.' The ITS changes b are not .in conformance with the letter and TSTF 52 as modified by staff comments. In particular, Amendments 173 and 155 for McGuire Units 1 and 2 respectively and Amendme~nts-144 and 138 for Catawba Units 1 and 2 respectively only approved 10 CPR 50 Appendix J Option for the Type A tests only. The. Type B and C tests must still be done in accordance L with Option A.7 Thus, only those leakage tests associated with Option B Type A test may be

? relocated to the Containment Leakage Rate Testing Program. This includes CTS 3.6.1.2.a.

1 CTS 3.6.1.2 ACTION a, 4.6.1.2,4.6.1.2.c and 4.6.1.2.1 with regards to Type A tests only. All

other CTS requirements specified above including CTS 4.6.1.2.1 must be retained in the ITS Eas SRs or Notes to the SRs. Comment: Licensee to update submittal with regards to 11/2/95

' letter, TSTF 52 as modified by staff comments and the above comments or provide additional justification for deviations.

DEC Response:

CTS 4.6.1.1.c is reduadant to the requirements in 10 CFR 50, Appendix J, Option A, Ill.D.2.

a DOC L.33 is addec: to justify the deletion of this detail. ITS SR 3.6.1.1 is revised to address Type' A testing and inspections and new ITS SR 3.6.1.2 is added to capture the Type B and C testing consistent with CTS 3.6.1.2.b,4.6.1.2.d. 4.6.1.2.d.2 (McGuire),4.6.1.2.d.3 (Catawba),

4.6.1.2.d.4 (McGuire),L .4.6.1.2.h,4.6.1.2.1 (Catawba), and 4.6.1.2.J (McGuire). CTS 4.6.1.2.1

. for McGuire is relocated to the Bases. CTS 4.6.1.2,4.6.1.2.c, and 4.6.1.6 are part of the scope. of the Type A testing performed pursuant to Option B and are not changed. Discussion of Changes A.3 and LA.1 have been revised accordingly. ITS 5.5.2 is also revised to match.

mc4_cr_3.6 - 3.6-2 ' July 2, 1998

?

)

1 McGuire & Catauba' Improved TS Review Comments-i

, 'ITS Section 3.6 . Containment Systems REVISED RESPONSE:

During the comment resolution meeting with NRC June 17,1998, a number of consistency issues were identified which are discussed below, i

j"

1. - (McGuire Only) The CTS markup and STS markup for SR 3.6.1.2 were not consistent-with tho' typed ITS. Note 2 was missing from the _STS markup and the second paragraph of the surveillance was missing from the CTS markup.. These markups have been corrected.

~

. 2. (McGuire Only) The CTS markup for CTS 4.6.1.2.d.4 and 4.6.1.2.1 are relocated moved to the Bases and the changes justified by LA1. DOC LA1 only describes changes moved to the Containment Leakage Rate Testing Program, not the Bases. New DOC .

LA28 has been added to justify the movement of this detail to the Bases and LA1 has been revised accordingly.

3. ' References to airlock testing and exemptions not being required in TS should be deleted from DOC LA1. DOC LA1 has been revised to delete these discussions.

i ' 4. (McGuire'Only)'The STS markup and CTS markup for SR 3.6.1.1 are inconsistent.

The reference to airlock testing has been deleted from the STS markup.

, . 5; Bases references _to 10 CFR 50, Appendix J, on pages B 3.6-1 does not specify the d applicable Appendix J option.: Tl.is statement is revised to include option B for SR

' 3.6.1.1 and add option A for SR 3.6.1.2.

6. Bases references in SR 3.6.1.1 to Type A acceptance criteria includes a phrase "following an outage or shutdown that included Type A testing." The quoted phrase -

has been deleted consistent with NRC comments on TSTF-52. Also the *and/or" in the

[ Bases for SR 3.6.1.1 and 3.6.1.2 are replaced with "and" consistent with the STS. l.

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2 mc4_cr_3.6 3.6-3 July 2, 1998

. . ;w , \

Containment B 3.6.1 B 3.61 CONTAINMENT SYSTEMS B 3.6.1:. Containment BASES

.a BACKGROUND The containment is a free standing steel pressure vessel surrounded by a reinforced' concrete reactor building. The containment vessel, including all its penetrations,.is a low leakage: steel she11' designed to contain the radioactive material. that may be released from the reactor core following a design basis Loss of Coolant Accident-(LOCA).

Additionally, the containment' vessel and reactor building provide shielding from the fission products that may be present in the containment atmosphere following accident; conditions.

The containment' vessel is a vertical cylindrical steel pressure vessel with hemispherical ' dome and a flat circular

' base. It:is completely enclosed .by a reinforced concrete reactor building. . An annular space exists between the walls and domes of the steel containment vessel and the concrete t(t reactor building to provide for the collection, mixing, holdup, and controlled release of containment out leakage.

Ice condenser containments' utilize an outer: concrete building for, shielding 'and an inner steel containment for O ' leak-tightness.

s Containment piping

' passage of process,service, penetration' sampling, assemblies provide' for the and instrumentation-pipelines into the containment vessel while maintaining containment. integrity. The reactor building provides shielding)and allows controlled release of the annulus atmosphere under accident conditions, as well as environmental missile protection for the containment vessel 7

and Nuclear Steam Supply System.

~

The inner steel containment and its penetrations establish the leakage limiting. boundary of the containment.

Maintaining the containment OPERABLE limits the leakage of fission product. radioactivity from the containment to the

= environment. SR 3.6.1.1 leakage rate requirements comply with 10 CFR 50, Appendix J ' Option B (Ref. 1), as modified by approved exemptions. SR 3.6.1.2 leakage rate :

requirements comply witt 4 CFR 50, Appendix J. Option A-7 (Ref. 1), as-modified by approved exemptions.

I Theitsolation devices for the penetrations in the.

, containment boundary are a part of the containment leak-(continued)

McGuire Unit 1 .B 3.6-1 Supplement 6- l-u m .. .

Containment B 3.6.1 f]

v BASES (continued)-

. ACTIONS Ad I In the event containment is inoperable, 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 Completion Time provides a period of time to correct the problem commensurate with the importance of maintaining containment OPERABLE during MODES 1, 2, 3,' and 4. This time period also ensures that the probability of an accident (requiring containment OPERABILITY) occurring during periods when containment is inoperable is minimal..

B.1 and 8.2 If containment.cannot be restored to OPERABLE status within the required Completion Time, 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 i 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed j Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without n challenging plant systems. j U f i

SURVEILLANCE SR 3.6.1.1 j REQUIREMENTS i Maintaining the containment OPERABLE requires compliance I with the visual examinations and Type A leakage rate test requirements of the Containment Leakage Rate Testing Program. Failure to meet specific leakage limits for the air lock, secondary containment bypass leakage path, and purge valve with resilient seals (as specified in LC0 3.6.2 and LC0 3.6.3) does not invalidate the acceptability of the overall containment leakage determinations unless the l specific leakage contribution to overall Type A, B, and C leakage causes one of these overall leakage limits to be exceeded. As left leakage prior to the first startup after performing a required Containment Leakage Rate Testing for overall Program leakage Type A leakage test is required following to or an outage beshutdown

< 0.75 L, that included Type A testing. At all other times between required leakage l

(~ (continued)

( !

i

-B 3.6-4 Sup; ement 6 l McGuire Unit 1 .

/

i

Containment B 3.6.1

.O V

BASES SURVEILLANCE SR 3.6.1.1 (continued) l REQUIREMENTS rate tests, the acceptance criteria is based on an overall Type A leakage limit of s 1.0 L,. At c 1.0 L the offsite dose consequences are bounded by the assump' ,ons of the safety analysis. SR Frequencies are as required by the Containment Leakage Rate Testing Program. These periodic testing requirements verify that the containment leakage rate does not exceed the leakage rate assumed in the safety analysis.

The Surveillance is modified by a Note which requires that the space between each dual-ply bellows assembly on containment penetrations between the containment building and the annulus. be vented to the annulus during each Type A test.

SR 3.6.1.2 Maintaining the Containment OPERABLE requires compliance with the Type B and C leakage rate test requirements of x 10 CFR 50, Appendix J, Option A (Ref.1), as modified by 4-) approved exemptions. Failure to meet specific leakage limits for the air lock, secondary containment bypass leakage path, and purge valve with resilient seals as I specified in LC0 3.6.2 and LCO 3.6.3 does not invalidate the acceptability of the overall containment leakage determinations unless the specific leakage contribution to Type A, B and C leakage causes one of these overall leakage l limits to be exceeded. - As left leakage prior to the first startup after performing a required 10 CFR 50, Appendix J, Option A, leakage test is required to be < 0.6 L, for combined Type B and C leakage. At all other times between required leakage rate tests, the acceptance criteria is based on an overall Type A leakage limit of s 1.0 L,. At s 1.0 L, the offsite dose consequences are bounded by the

. assumptions of the safety analysis. SR Frequencies are as required by Appendix J, Option A, as modified by approved exemptions. Thus, SR 3.0.2 (which allows Frequency extensions) does not apply. These periodic testing ,

requirements verify that the containment leakage rate does j not exceed the leakage rate assumed in the safety analysis. '

-( (continued)

(

.McGuire Unit 1 .B 3.6-5 Supplement 6 l

i Containment B 3.6.1

/~T B 3.6 CONTAINMENT SYSTEMS B 3.6.1 Containment i

BASES BACKGROUND The containment is a free standing steel pressure vessel surrounded by a reinforced concrete reactor building. The containment vessel, including all its penetrations, is a low u

leakage steel shell designed to contain the radioactive material that may be released from the reactor core following a design basis Loss of Coolant Accident (LOCA).

Additionally, the containment vessel and reactor building provide shielding from the fission products that may be present in the containment atmosphere following accident conditions.

1 The containment' vessel is a vertical cylindrical steel pressure vessel with hemispherical dome and a flat circular l base. It is completely enclosed by a reinforced concrete reactor building. An annular space exists between the walls and domes of the steel containment vessel and the concrete reactor building to provide for the collection, mixing, holdup, and controlled release of containment out leakage.

Ice condenser containments utilize an outer concrete l (

building for shielding and an inner steel containment for

( leak tightness.

Containment piping penetration assemblies provide for th9 passage of process, service, sampling, and instrumentation pipelines into the containment vessel while maintaining containment integrity. The reactor building provides shielding and allows controlled release of the annulus atmosphere under accident conditions, as well as environmental missile protection for the containment vessel and Nuclear $ team Supply System.

The inner steel containment and its penetrations establish the leakage limiting boundary of the containment.

Maintaining the containment OPERABLE limits the leakage of fission product radioactivity from the containment to the envi ronment. SR 3.6.1.1 leakage rate requirements comply with 10 CFR 50, Appendix J. Option B (Ref.1), as modified by approved exemptions. SR 3.6.1.2 leakage rate requirements comply with 10 CFR 50, Appendix J, Option A ;

l (Ref.1), as modified by approved exemptions.

.The isolation devices for the penetrations in the containment boundary are a part of the containment leak (continued)

-(GT McGuire Unit 2 .8 3.6 Supplement 6 l l

w__-_________

Containment B 3.6.1 BASES (continued)

ACTIONS AJ I In the event containment is inoperable, 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 Completion Time provides a period of time to correct the problem commensurate with the importance of maintaining containment OPERABLE during MODES 1, 2, 3, and 4. This time (

period also ensures that the probability of an accident (requiring containment OPERABILITY) occurring during periods when containment is inoperable is minimal.

B.1 and 8.2 If. containment cannot be restored to OPERABLE status within the required Completion Time, 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 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without i g challenging plant systems. j t

SURVEILLANCE SR 3.6.1.1 REQUIREMENTS Maintaining the containment OPERABLE requires compliance with the visual examinations and Type A leakage rate test !

requirements of the Containment Leakage Rate Testing Program. Failure to meet specific leakage limits for the air lock, secondary containment bypass leakage path, and purge valve with resilient seals (as specified in LC0 3.6.2 and LC0 3.6.3) does not invalidate the acceptability of the overall containment leakage determinations unless the l specific leakage contribution to overall. Type A, B, and C leakage causes one of these overall leakage limits to be ;

exceeded. As left leakage prior to the first startup after '

performing a required Containment Leakage Rate Testing Program leakage test is required to be < 0.75 L, for overall Type A leakage following an outage or shutdown that included i Type A testing. At all other times between required leakage 1

I I

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T (continued) ,

I McGuire Unit 2 .B 3.6-4 Supplement 6 !

l

Containment B 3.6.1.

BASES SURVEILLANCE SR 3.6.1.1 (continued) 'I REQUIREMENTS rate tests, the . acceptance criteria is- based on an overall j Type A-leakage limit-of s 1.0 L . At s 1.0' L the offsite j dose consequences are bounded by the. assumpti,ons of the -

safety analysis. SR Frequencies are as required by the Containment Leakage' Rate Testing Program. These periodic ,

testing requirements verify that the containn;ent leakage {

rate does not exceed the leakage rate assumed in the safety analysis.

1 The Surveillance is modified by 'a Note which requires that

, the space between each dual-ply bellows assembly on {

containment penetrations between the containment building and the annulus be vented to the annulus during each Type A' ;

test.

.SR 3.6.1.2 Maintaining the Containment OPERABLE requires compliance with the Type B and C leakage rate test requirements of

( 10.CFR 50, Appendix J, Option A (Ref. 1), as modified by approved' exemptions. Failure to meet specific leakage limits for the air lock,' secondary containment bypass leakage path, and purge valve with resilient seals as.

specified in LCO' 3.6.2 and LCO 3.6.3 does not invalidate the acceptability of the overall containment leakage determinations unless the specific leakage contribution to Type A, B and C leakage causes one of these overall leakage- j limits to be exceeded. As left leakage prior to the first startup after performing a required 10 CFR 50, Appendix J.

Option A, leakage test is required to be < 0.6 L, for

! combined Type B and C leakage. At all other times between required leakage rate tests, the. acceptance criteria is based on an overall Type A leakage limit of s 1.0 L.. At s 1.0 L, the offsite dose consequences are bounded by the assumptions of the safety analysis. SR Frequencies are as required by Appendix J, Option A, as modified by approved exemptions. Thus, SR 3.0.2 (which allows Frequency extensions) does not apply. These periodic testing requirements verify that the containment leakage rate does not exceed the leakage rate assumed in the safety analysis.

. 1

)

(continued)

McGuire Unit 2 .B 3.6-5 Supplement 6

/

-s \ Specification 3.6. l Y)

INSERT la ,

SR 3.6.1.1 -------------------NOTE-------------------

The space between each dual ply bellows assembly on penetrations between the containment building and annulus shall be vented to the annulus during Type A tests.

In accordance with Perfonn required visual examinations and Type A the Containment leakage rate testing in accordance with the Leakage Rate Containment Leakage Rate Testing Program. Testing Program SR 3.6.1.2 -------------------NOTE--------------------

1. Following each Type A test, the space between each dual-ply bellows assembly shall be subjected to a low pressure test at 3 to 5 psig to verify no detectable leakage, or the assembly shall be subjected to a leak test with the pressure on the containment side of the assembly at Pa-

) 2. Type C tests on penetrations M372 and M373 may V be performed without draining the glycol-water

NOTE------

mixture from the seats of their diaphragm valves if meeting a zero indicated leakage rate (not SR 3.0.2 is not including instrument error). applicable

......s...................................

Perform required Type B and C leakage rate testing, except for containment air lock testing and valves In accordance with with resilient seals, in accordance with 10 CFR 50, 10 CFR 50, Appendix J Option A, as modified by approved Appendix J, Option exemptions. A, as modified by approved exemptions The leakage rate acceptance criterion is S 1.0 Lg.

However, during the first unit startup following testing performed in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions, the leakage rate acceptance criteria are

< 0. 6 La for the Type B and Type C tests.

McGuireI Page 2 of (,

1 L______

i i

5f L'ista6 E.&./

T s TEMS

$ EILL REOUIREM S (Co inuedi.

Deleted

. I *

b. Dele l

c. The accuracy of each test in accordance wi A test shall be verified by a su N

Reeulatory Guide 1.163, September,p[ementa 1995.

d. ,

50.54(o) and 10 CFR 50 Appendix J, Type B and C tests sh tion A, with gas at P ,

14.8 ig, at intervals no greater invol ng: n 24 months except for tests s t2 2, b d*'l "

i MirJ6clD

@Q:wam='M,::ere"**-o Q) s-nd epust isolp(fon valves /with resfifent) nog 4, 4) Ct ts perio s a. 3. t,. l. on contal penetrat s M372, '73 raining 1 1-water xture f their aphrage val NF-228A, 2338, a seats f yy meeti a zero ind -234A) f L -

_ erme far the d' leakage te (not i luding trument, ragevalves.gThesetes may be )

li f tests wh' are otne requi es in 1

)

.2(a) of 1 FR 50. Appe Secti a.n x J to us air or a regen asj ,

O J

i i

l l

l i

McGUIRE - int!T 1 . 3/4 6-3 i Amendment No.173 O "s* & \

l

( #

)

Qr:.k t$1b; 9,4.,)

CONT d TEMS Elf f RE EMENT (Conti dl /

yi n

rg sea ska 1>be and demons o Specificati ted OP the requi

.6.1.g.3 or .1.g.4, as e.,. able; See Nu [Thecombinedbypassleakagerateshallbedeterminedtobele

\

t+ IT53. 3 . 0.07 L by appitcable Type a and c tests at least once per '

except,for penetrations which are not individually e; testabl. 24 mont no detectable leakage when tested with soap bubb i

, Q is pressurized to p.,14.8 psig during each Type A test;

{

cat 1 pkg " h. ~ The s between each dual-ply bel m,q,).) une assembly on containment

{ .w ions between the contai building and the annulus sha l p o esca ed ryatoathe annulus durineJ l sest A tests the spac w _;;;. eacn c.ual-ply r ro11ovin compte be lows ssembly 5 ull be su,b ected veri no detectable e a low pressure test at 3-5 g to or the dual-ply bellows ass y shall be s jected to a leak e t with the prirssure on the e side of the dual inment

],g,1. leakage to be vi lows assembly at Ifnits of '+eciff .14.8 psi to verify the l i

tion 4. 1.2f.:

1. All test teaka

. converted to ratesvalues.

solute sliall be calculated Error analysesusin11 observed data a t a ha need lategrated Leakage Measu be' performed to nt System; Md 4 .

Sase ss.e ##I': b. The provision 5 Specification 4.0.2 are not applicable.

] I Nek. b \

SR.T.G.\.t.-

4 i

McGUlitE - IMIT 1 3/4 6-4 l

Amendment No.173 fAf SD h

?

L___-____-_____ __

i Specification 3.6. f INSERT la ;

SR 3.6.1.1 -------------------NOTE-------------------

l The space between each dual ply bellows assembly on penetrations between the containment building and annulus shall be vented to the annulus during Type A tests.

In accordance with Perform required visual examinations and Type A the Containment leakage rate testing in accordance with the Leakage Rate Containment Leakage Rate Testing Program. Testing Program SR 3.6.1.2 -------------------NOTE--------------------

1. Following each Type A test, the space between each dual-ply bellows assembly shall be subjected to a low press"re test at 3 to 5 psig to verify no detectable nakage, or the assembly shall be subjected to a leak test with the pressure on the containment side of the assembly at Pa-O 2. Type C tests on penetrations H372 and M373 may be performed without draining the glycol-water mixture from the seats of their diaphragm valves ~~~~~'

k,-~~~~~

if meeting a zero indicated leakage rate (not SR 3.0,2 is not including instrument error). applicable Perform required Type B and C leakage rate testing, except for containment air lock testing and valves In accordance wit,a with resilient seals, in accordance with 10 CFR 50, 10 CFR 50, Appendix J, Option A, as modified by approved Appendix J, Option exemptions. A, as modified by approved exemptions The leakage rate acceptance criterion is S 1.0 La.

However, during the first unit startup following testing performed in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions, the leakoge rate acceptance criteria are

< 0. 6 La for the Type B and Type C tests.

McGuire 1 Page2.of4 s

snnas

[CONTAI_ SYSTEMS s.s.t ,

O- Ae)

\SURVE E REQUIR TS (Con nued) '

l

a. Delete i

.I' b. Del g ,

g

$ E witI NryY NM D SR MA 1 t

C 50.54(o) 14.8 psig and 10 CFR 50 Appendix J. Option A, w

[nvolving:, at latervals no greater than 24 months except [or tests Afr M dfI th a a exhaust Iseptionvalvegirithres/te@ ,

!' [Ng3 y,t1. 4 Type C ts performed on tainment penetrati without raining the 31 s 10 72, 10 7 their aphrage valves 1-water aiutare f the seats of )

seet -228A,llF-2338 llF-234A), if sTrr i e.

a zero indica leakage ra+= faa,+

1 va- +h. d F " ~1= dine tesi. '

valvee_ IThese sts any be 11'eu of tests whi are othenrise n ~

Ogg 3.Q.C.2(a)of1, 50, Appendix J requ by Section use air or att 7 e asj

) 1' j

y ,

e MI - WIT 2 3/4 6-3 Amendment alo.155 l

0 ,

MF V'S fo e

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

Sp M l rat //in 5.C./

fCONTAlmudrSYST O. Ld :/ a-" .h ,

rv tc th e-4./ . pu 1 se' s 1 ted and t is atton valves. th resiti materi strated by L Specifi , on 4.6.1.9 or 4.6.1.9 . as appilcable;requi ff. __

See /1/F4' The O.07 combined bypass leakage rate shall be determined to be less than by applicable Type 8 and C tests at least once per 24 monAs

' of / 75 % .3 except(for penetrations dich are. net individually te

. no detectable leakage When tested with soap bubbl (tainment is pressortred to p.,14.8 psig during each Type A testi j 5 . 1 m is f* ( h. The s between each 1-ply bellows ssembly on cent ament S M 6*ll tions between containment 1 ding and the ted to the us during Type les shall

4 of <== yype A ses a embly shall be ubjected to a 1 pressure sne space De tests. rFoll each dual-ply completten ~

lows rify no de test 3-5 esig to

!- le leakage or dual-ply bell

! ' cubjectedto leak test with assembly shall '

side of the pressere on

-ply bellows as containment

[,$,g,t leakage to be thin the limit of Specifi., ly at p psig to verify the jn 14 4.6.1.2f.;

cati l V A. All test lan e 4 .1 % s :ge rates shall

,solete value . Error anal calculated as observed data convertedflanced!!itegrat i

ap Leakage Meas shall be performed'to )

System; and

j. /

i fe A'kksf e The provisions of st+;ification 4.0.z are not applicaby i

ys f,S.A ,

N61er W N SR 3.t .b1 w

(

n McGUIRE - WIIT 2 3/4 5-4 Amendment No.155 t

l pq 5 $ &

P

Discussien of Chang:s Secticn 3.6 - C:ntainment Systems TECHNICAL CHANGES - REMOVAL OF DETAILS LA.1 CTS Surveillance Requirements 4.6.1.2.c,

and 4.6.1.6 contain details and czemption; for meeting the leak rate testing and visual inspection requirements stated in 10 CFR 50, Appendix J Option 8 for Type A testing. ITS SR 3.6.1.1 contains the broader requirement that all applicable Type A testing and inspection specified by 10 CFR 50, Appendix J, Option B must be met. ITS 3.5.2 include; the requirc=nt; for the

, centcinn nt air lock;. The specific acceptance criteria for Type A 1eakage rate testing and requirement to comply with RG 1.163 are retained _in ITS 5.5.2. The details of how the requirements of 10 CFR 50 Appendix J. Option B are met are relocated to the Containment Leakage Rate Testing Program. Descriptive details related to program implementationnpprcued czc=ptions are not necessary for inclusion within the TS. Changes to this program will be evaluated under the site procedure control program to ensure compliance with the CFR and approved exemptions. This change is consistent with the NUREG-1431 philosophy of relocating certain details outside of the TS.

h V

LA.2 The descriptive information in the CTS LC0 3.6.1.3.a. regarding OPERABILITY of the air locks, is being moved to the Bases for ITS l

3.6.2. The movement of this information is appropriate because it involves details that are not necessary for inclusion in the LCOs and are more appropriate for the Bases. The Bases are subject to j the controls described in ITS Chapter 5 " Administrative Controls."

Changes to the Bases are evaluated under the 10 CFR 50.59 criteria. Any change, using this criteria, will ensure proper review. This change in ITS 3.6.2 is consistent with NUREG-1431.

LA.3 CTS Surveillance 4.6.1.3.a and b contain detailed information describing the airlock testing required by 10 CFR 50, Appendix J.

This descriptive information is moved to the Bases and replaced with the ITS SR 3.6.2.1 which references 10 CFR 50, Appendix J, Option A. The frequency _of testing is required by 10 CFR 50, Appendix J, Option A and the acceptance criteria are retained in

.ITS SR 3.6.2.1. Other detail information contained in the CTS Surveillance, including any exemptions to 10 CFR 50, Appendix J, is relocated to the Bases. The movement of this information is !

i appropriate because it involves details that are not necesst.ty for )

inclusion in the LCOs and are more appropriate for the Bases. Any

,(' changes to the Bases are controlled by the Bases control

>\ l l

lMcGuireUnits1and2 Pdge LA - 18 Supplement 62 l

1 I

< a

Discussicn cf Charg2s S:ctica 3.6 - Containment Systems Q TECHNICAL CHANGES - REMOVAL OF DETAILS LA.1 CTS Surveillance Requirements 4.6.1.2.c, " ' ' " ^ I" ' ' 4 and 4.6.1.6 contain details and exemptfe-for meeting the leak rate testing and visual inspection requirements stated in 10 CFR 50, Appendix J, Option B for Type A testing. ITS SR 3.6.1.1 contains the broader requirement that all applicable Type A testing and inspection specified by 10 CFR 50, Appendix J, Option B must be met. !!S 2.5.2 include; the requirement; for the contain=nt air lock;. The specific acceptance criteria for Type A 1eakage rate testing and requirement to comply with RG 1.163 are retained in ITS 5.5.2. The details of how the requirements of 10 CFR 50 Appendix J, Option B are met are relocated to the Containment Leakage Rate Testing Program. Descriptive details related to program implementationApproved exempticn; are not necessary for inclusion within the TS. Changes to this program will be evaluated under the site procedure control program to ensure compliance with the CFR and approved exemptions. This change is consistent with the NUREG-1431 philosophy of relocating certain details outside of the TS. j i

( LA.2 The descriptive information in the CTS LC0 3.6.1.3.a. regarding C OPERABILITY of the air locks, is being moved to the Bases for ITS 3.6.2. The movement of this information is appropriate because it involves details that are not necessary for inclusion in the LCOs j and are more appropriate for the Bases. The Bases are subject to I the controls described in ITS Chapter 5 " Administrative Controls." l Changes to the Bases are evaluated under the 10 CFR 50.59 criteria. Any change, using this criteria, will ensure proper review. This change in ITS 3.6.2 is consistent with NUREG-1431.

LA.3 CTS Surveillance 4.6.1.3.a and b contain detailed information I describing the airlock testing required by 10 CFR 50, Appendix J.

This descriptive information is moved to the Bases and replaced with the ITS SR 3.6.2.1 which references 10 CFR 50, Appendix J, Option A. The frequency of testing is required by 10 CFR 50, Appendix J, Option A and the acceptance criteria are retained in ITS SR 3.6.2.1. Other detail information contained in the CTS Surveillance, including any exemptions to 10 CFR 50, Appendix J, l

is relocated to the Bases. The movement of this information is appropriate because it involves details that are not necessary for inclusion in the LCOs and are more appropriate for the Bases. Any

(] changes to the Bases are controlled by the Bases control V

I j McGuire Units 1 and 2 P ge LA - 18 Supplement 6B h

j

Discussicn of Ch ng:s S;ction 3.6 - Containment Systems l 'b v TECHNICAL CHANGES - REMOVAL 0F DETAILS described in ITS Chapter.5 " Administrative Controls."5 Changes to the Bases are evaluated under the 10 CFR 50.59 criteria. Any change, using this criteria, will ensure proper review. These changes are consistent with NUREG-1431.

LA.27 CTS 1.27 items a and c define the attributes of reactor building operability and integrity. These attributes have been relocated

, to the Bases for ITS 3.6.16. The descriptive attributes are more appropriate information for Bases and are not necessary to be included within tM Technical Specification. The requirement to maintain an operable reactor building isretained in ITS 3.6.16.

The Bases are subject to the controls described in Chapter 5

" Administrative Controls" of the ITS specifications. Changes to the Bases are evaluated under the 10 CFR 50.59 criteria. Any change, using this criteria, will ensure proper review. This change is consistent with NUP.EG-1431.

LA.28 The descriptive information in the CTS 4.6.1.2.d.4 and 4.6.1.2. i regarding descriptive leakage rate testing methods and details is

O being moved to the Bases for ITS 3.6.1. The movement of this information is appropriate because it involves descriptive details that are not necessary for inclusion in the LCOs and are more f appropriate for the Bases. The requirement to conduct leakage rate testing is retained in ITS SR 3.6.1.1 and 3.6.1.2. The Bases are subject to the controls described in ITS Chapter 5

" Administrative Controls. " Changes to the Bases are evaluated

under the 10 CFR 50.59 criteria. Any change, using this criteria, will ensure proper review. This change is consistent with NUREG-1431. \

i l

l' I

s d

I McGuire Units 1 and 2 Page LA.- 88 Supplement 63 l

)

I Containment (EiiRLr k. Li-~-- ica Condenser. nM rh=n

'~~~~

~

- - - . - -- -- l h sp.a st4*c u <a A t*~skh% LeW se - e 4.n. F yN, 9.

ben"--l. ply $cbs y a4 m n aa.f.a SURVEILLANCE REQUIREMENTS 8 b il 3e @ _4e d 4. N u a M 1 4re- O p _ , J<y b . ,

)

~ -_ - ,

~

f SURVL1LU n t ~ FREQUENCY SR 3.6.1.1 Perform required visual examinations and C---- TE----

! Q J 1eakage rate testingei!xcept s con u... - m SR .0.2 is t t

O '" g l

. Sir sorr testantain accoraance with . licable ,446 l

{0 v d J. as moa ~ied un ouexemptions.

em

-}14/c pprov fq $

l O(p The eakage rate accepta e criterion is. In accordance M"J s .0 L . However, dur' g the first uni with u u n artup following tes ng performed in ix J e [

i accordance with 10 bo, Appendix J as ified ,

} Imodified by approv exenptions, the prove j l

leakage rate ace tance criteria ar < 0. x -nt' n for the Type l and Type C test and 1

L,0.75

< L, for Type A test.

L ust%TM

_ - - 2 SR 3.6.1 Verify containment ructural integrity In ecordance \

i

't V)

(N in accordance wi Surveillance Pro am.

the Containment Tendon w' h the ontainment

)3 endon j / Surveillance Program

/ ~

l 1

i i !

I i

WOG STS 3.6-2 Rev 1, 04/07/95 l 4eb 1 '

l I

<_.-___--.______ - - _ . J

INSERT V

SR 3.6.1.2 -----------------NOTES----------------- , i 1.- Following each Type A test, the

I space between each dual-ply bellows assembly shall be subjected to a low pressure test at 3 to 5 psig to verify no detectable leakage, or the assembly shall be subjected to a leak test with the pressure on the containment side of the assembly at P .

2. Type C tests on penetrations M372 i and M373 may be performed without I draining the glycol-water mixture i from the seats.of their diaphragm i valves if meeting a zero I indicated leakage rate (not ---- NOTE------

including instrument error). SR 3.0.2 is not i

_________....... __.._____.._____....- applicable Perform required Type B and C leakage rate testing, except for containment In accordance with n air lock testing and valves with 10 CFR 50,

{

resilient seals, in accordance with Appendix J, Option (V) 10 CFR 50, Appendix J, Option A, as A, as modified by ,

modified by approved exemptions. approved exemptions The leakage rate acceptance criterion i is s 1.0 L,. However, during the first unit startup following testing ]

performed in accordance with 10 CFR 50, l Appendix J, Option A, as modified by approved exemptions, the leakage rate ]

j acceptance criteria are < 0.6 L, for the Type B and Type C tests.

.g INSERT Page 3.6-2 3 McGuire 1

1

~

O' .

Containment q3.6.1N%

B 3.6 CONTAllMENT SYSTEMS l B 3.6.1 Containment 5cet,oosenser]

l l- BASES

. (f) l . 6mE)

- BACKGROLDO Thecontainmentisafreestanding(steelpressurevessel surrounded by a reinforced concrettiketEDdD building. .The i containment vessel, including all its penetrations, is a low g L leakage steel shell designed to contain the radioactive y,,

l t material that may be relaa=ad from t'a reactor enre followine a si sesis[Accidentv e u.

m

"" - >- containment; building prov ce snielding Additiana11v.

from the the M " O W tission products that may be present in the containment

[vesse@ { atmosphere following accident conditions.

O The containment vessel is a vertical cylindrical steel b

4 pressure vessel with hemispherical dome and a base.

r = = =r a-i --=nn. It is completely enclosed by a trac fe.- reintorced concreti!;te2EEZD building. An annular space exists between the walls and domes of the steel containment vessel and the concrete shield building to provide for the holdu O collection, mixing,kaoe.

containment out lea p, and

-Ice controlled condenser release of utilize containments an outer concrete building for shielding and an inner steel contairunent for leak tightness.

Containment piping penetration assemblies provide for the passage of process, service, sampling and instrumentation

. pipelines into the containment vessel,while maintaining gr4 M containment integrity. The stitallkbuilding >rovices

~ -

shielding and allows controlled reTease of tw annulus O-atmosphere'under accident conditions, as well as environmental missile protection for the containment vessel and Nuclear Steam Supply System.

The inner steel conta hment and its penetrations establish the leakage limiting boundary of the containment.

Maintaining the containment OPERABLE limits the leakage of g ,g,t.l - fission product radioactivity from the containment to the environment.

A t.r.Aor e, M with 10 CFR 50,SR 3.6.1.1 leakab Appendix Re rate 1), asrequirements conply modified by approved C.mpl y sdW MM.m exenptions.j gg N The isolation' devices for the penetrations in the W k e.* m*MN. containment boundary are a part of the containment leak y app m J h r tight barrier. To maintain this leak tight barrier:

(continued) l WOG STS B 3.6-1 Rev 1, 04/07/95 Q

l v. cmA M-9

q V.

Containment n ce ron s m5r$

B 3.6.1 BASES (continued)

ACTIONS Al In the event containment is inoperable, cor.tainment 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 Completion Time provides a period of time to correct the problem consensurate with the importance of maintaining containment OPERABLE during HDDES 1, 2. 3, and 4. This time period also ensures that the probability of an accident j (requiring containment OPERABILITY) occurring during periods when containment is inoperable is minimal.

B.1 and B.2 If containment cannot be restored to OPERABLE status within the required Completion Time, 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 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full (3) power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.1.1 REQUIREMENTS TW A Maintaining the containment OPERABL requires compliance with the visual examinations and_ eakage rate test __

requirement.s ofAjo ON bo Annenm x o t kau ii at Q- -

- ov enrnvm mmnti6ns J Failure to meetrair lock 1, e u n u n> r' .

'N k O F-(bc Orrt/wf*t<M secon'dary containment bypass leakage path m and ource valve 8'~^ k no 1

(e,g'M&Jc,1**Sdl/ LandLC03.6.3Mdoesnotinvalidatetheacceptabilityof with resilient seall'pakmWerspecified in LCO@^** 3.6.2]sS (,54lb P'# theCB overallpeakage determinations unless thedb -

contribution to eM W D Type A, B, and(C leakaoe caus Gi

(* uo exemVHmith. As left leakage prior to the fir startuo after performing a reauired n ruH 50f_ ADDendX1i W'Yl --

(G,,4;n w ,dh red to beldh dr%ihedAypaJ __ft # *L M t,4 Bal< l lenkage test is

"'^=L 0.75 L, for overall Type A leaka YM'" W% l l

G @/,"4 P<ce all'othertide' between required leakage rate tests, t t

acceptance criteria is based on an overall Type A leakage 4"<v; d a

, llwh limit of s 1.0 L,. At s 1.0 L the offsite dose ,. Jag ar- l xJ tu- consequences are bounded by th,e assumtions of the safety sxMawo 1 l ,d ,, analysis. SR Frequencies are as required by,Ormmrow SJ ,uli.Jed j r$/- f

((y,c (o

(Ae G os k m e, & Lea t a ble Teshony Prognues.q (continued)

T'!A' A -

'M l

l (q/

WOG STS B 3.6-4 Rev 1, 04/07/95 (VcLtJa.>)

9 L___ _

INSERT LJ The Surveillance is modified by a Note which requires that the space between each dual-ply bellows asseculy on containment penetrations betweeri the containment building and the annulus be vented to the annulus during each Type A test.

SR 3.6.1,2 Maintaining the Containment OPERABLE requires compliance with the Type B and C leakage rate test requirements of 10 CFR 50, Appendix J Option A (Ref. 1), as modified by approved exemptions. Failure to meet specific leakage limits for the air lock, secondary containment bypass leakage path, and purge valve with resilient seals as specified in LC0 3.6.2 and LCO 3.6.3 does not invalidate the acceptability of the overall containment leakage determinations unkss the F specific leakage contribution to Type A, B and C leakage causes one of tPese overall leakage limits to be exceeded. As left leakage prior to the first startup after performing a required 10 CFR 50, Appendix J, Option A, leakage test is required to be < 0.6 L for combined Type B and C leakage. At all other times between required el,akage rate tests, the acceptance criteria is based on an overall Type A leakage limit of s 1.0 L,. At s 1.0 L, the offsite dose consequences are bounded by the assumptions of the safety analysis. SR Frequencies are as required by Appendix J, Option A, as modified by approved exemptions. Thus, SR 3.0.2 (which allows Frequency extensions) does not p)

(

apply. These periodic testing requirements verify that the containment leakage rate does not exceed the leakage rate assumed in the safety analy: s.

The Surveillance is modified by two Notes. Note 1 requires that following 6dch Type A test, the space between each dual-ply bellows assembly be subjected to a low pressure leak test with no detectable leakage. Otherwise, the assembly must be tested with the containment side of the bellows assembly pressurized to P and meet the requirements of SR 3.6.3.8 (bypass it:akage requirements). bote2allowspenetrationsM372andM373tobetestedwithout

{

draining the glycol-water mixture from the associated diaphragm valves (NF- l 228A, NF-233B, and NF-234A) as long as no leakage is indicated. This test may be used in lieu of Section III.C.2(a) of 10 CFR 50, Appendix J Option A which requires air or nitrogen as the test medium. The required test pressure and interval are not changed.

l All test leakage rates shall be calculated using observed data converted to absolute values. Error analysis shall also be performed to select a balanced integrated leakage measurement system.

l INSERT Page B 3.6-5

'd McGuire

(

McGuire & Catawba Improved T5 Review Comments-ITS Section 3.6, Containment Systems

~%h) l 3.6.1 -3 . DOC A.5 (ITS 1.0)

l. DOC LA.1' JFD Bases 1 JFD Bases 2 (McGuire only)

CTS 1.7

- CTS 4.6.1.2.d.3) --

CTS 4.6.1.2.h STS B3.6.1 Bases BACKGROUND

, ITS B3.6.1 Bases BACKGROUND CTS 1.7 provides the definition for Containment Integrity and is justified by DOC A.5 as deleted in the CTS Markup of ITS 1.0. This is incorrect. The definition is part of the technical specifications and as such delineates CTS requirements. Therefore, an appropriate markup of CTS 1.7 should be included in the CTS Markup of ITS 3.6.1,3.6.2 and 3.6.3. In particular, all of CTS 1.7 (a through e, except d, is relocated to ITS B3.6.1 Bases - BACKGROUND.- This change would be considered as a Less Restrictive (LA) change, in addition, ITS B3.6.1 Bases BACKGROUND shows that STS B3.6.1 Bases - BACKGROUND item d has been

' deleted. . Based on CTS 1.7.e,4.6.1.2.d.3), and 4.6.1.2.h, this item cannot be deleted.

Furthermore, CTS 1.7a, c and a should also indicate that they are associated with other ITS 3.6.1,3.6.2, and 3.6.3 SRs which are Administrative changes. (See Comment Numbers O 3.6.2-1 and 3.6.3-1) Comment: Revise the CTS markup of ITS 3.6.1 to include a markup of CTS 1.7 and revise ITS B3.6.1 Bases BACKGROUND, accordingly. Provide additional discussion and justifications for the above Less Restrictive (LA) and Administrative changes.

DEC Responee:

' CTS 1.7 has been added to the CTS Markup for ITS Section 3.6. LA.25 has been added to the Discussion of Changes for CTS 3.6 to justify the movement of this information to the

.g Bases of ITS 3.6.1. CTS 1.7 item d is shown as being captured by the surveillance' requirement for ITS 3.6.1. No specific administrative changes beyond the general formatting l1 change (A.1) is necessary.

REVISED RESPONSE:

The CTS definition of containment' integrity states that the equipment hatches are closed and sealed. The iTS 3.6.1 Bases are revised to include "and sealed" consistent with the CTS.

(

-mc4_cr_3.6I 3.6-4 July 2, 1998 0,

. v

Containment B 3.6.1

(^}

LJ BASES BACKGROUND tight barrier. To maintain this leak tight barrier:

(continued)

a. All penetrations required to be closed during accident conditions are either:

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

2. closed by manual valves, blind flanges, or de-activated automatic valves secured in their closed positions, except as provided in LC0 3.6.3, " Containment Isolation Valve.s";

b. Each air lock is OPERABLE, except as provided in LC0 3.6.2,. " Containment Air Locks";

l c. All equipment hatches are closed and sealed; and

d. The sealing mechanism associated with a penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.

APPLICABLE The safety design basis for the containment is that the (V) SAFETY ANALYSES containment must withstand the pressures and temperatures of the limiting Design Basis Accident (DBA) without exceeding the design leakage rates.

The DBAs that result in a challenge to containment OPERABILITY from high pressures and temperatures are a loss of coolant accident (LOCA) and a steam line break (Ref. 2).

In addition, release of significant fission product radioactivity within containment can occur from a LOCA. In the DBA analyses, it is assumed that the containment is OPERABLE such that, for the DBAs involving release of fission product radioactivity, release to the environment is controlled by the rate of containment leakage. The containment was designed with an allowable leakage rate of 0.3% of containment air weight per day (Ref. 3). This leakage rate, used in the evaluation of offsite doses resulting from accidents, is defined in 10 CFR 50, Appendix J (Ref.1), as L,: the maximum allowable containment leakage rate at the calculated peak containment internal pressure (P,) resulting from the limiting design basis LOCA. The allowable leakage rate represented by L, forms the basis for the acceptance

(] (continued)

V l McGuire Unit 1 .B 3.6-2 Supplement 6 L_ ___ _ _ ._ . - -

Containment B 3.6.1 O BASES l V BACKGROUND tight barrier. To maintain this leak tight barrier:

(continued) l

a. All penetrations required to be closed during accident conditions.are either:

! 1. capable of being closed by an OPERABLE automatic containment isolation system, or

2. closed by manual valves, blind flanges, or de-activated automatic valves secured in their closed positions, except as provided in LC0 3.6.3, " Containment Isolation Valves";

b. Each air lock is OPERABLE, except as provided in LCO 3.6.2,. " Containment Air Locks";

l c. All equipment hatches are closed and sealed; and

d. The sealing mechanism associated with a penetration l (e.g., welds, bellows, or 0-rings) is OPERABLE.

(G APPLICABLE SAFETY ANALYSES The safety design basis for the containment is that the containment must withstand the pressures and temperatures of j the limiting Design Basis Accident (DBA) without exceeding the design leakage rates.

! The DBAs that result in a challenge to containment )

OPERABILITY from high pressures and temperatures are a loss of coolant accident (LOCA) and a steam line break (Ref. 2).

In addition, release of significant fission product radioactivity within containment can occur from a LOCA. In the DBA analyses, it is assumed that the containment is OPERABLE such that, for the DBAs involving release of fission product radioactivity, release to the environment is controlled by the rate of containment leakage. The containment was designed with an allowable leakage rate of 0.3% of containment air weight per day (Ref. 3). This leakage rate, used in the evaluation of offsite doses resulting from accidents, is defined in 10 CFR 50, Appendix J (Ref.1), as L,: the maximum allowable containment leakage rate at the calculated peak containment internal pressure (P,) resulting from the limiting design L basis LOCA. The allowable leak' age rate represented by L, forms the basis for the acceptance j f3 (continued) )

Q l f

l McGuire Unit 2 .B 3.6-2 Supplement 6 l

l

L i

l -

I T

\

,J.

Containment Ocefondens70 g g3.6.1 i

BES l BACKGROUND a. All penetrations required to be closed during accident l

(continued) conditions are either:

i 1. capaHe of being closed by an OPERABLE automatic con

Anment isolation system, or

2. closed by manual valves, blind flanges, or r

de-activated automatic valves secured in their closed positions, except as provided in LC0 3.6.3, " Containment Isolation Valves";

@@ 8"I'b b.

c.

Each a'- lock is OPERABLE, exce LC0 3.6.2,Contaiment Air Lockt All equipment hatches are closed R s*:

rovided in ahs l

I

/ gT u/cM 7 ,

i The(orehdi7adsalingmechanisma[sociatedwitha

! penetration is erablecexcesgemarpme in

_ (coq.6O. F /

vTe.3., u m , kit.-3, .e S Sh g

~ l

, ,' .M lVj APPLICABLE The safety design basis for the containment is that the t

SAFETY ANALYSES contaiment must withstand the pressures and temperatures of the limiting GDLj without exceeding the design leakage rates.

! Desdn Besi6 The DBAs that result in a challenge' to containment Amhnd(060 OPERABILITY from high pre res and tenperatur are a Q2 m cooient ace 50enu a steam line brea (Pip N un ACP Ment aMtAD( f. 2), jn add 1Llon, re east of significant fission product radioactivity within containment In the DBA analyses, it is i

can occur assumed thatfrom a LOCA @ t is OPERABLE such that, for the, the containmen l DBAs involving release of fission product radioactivity, g3 g L release to the environment is controlled by the rate of containment leakage. The containment was designed with an 4 J allowable leakage rate of . i containment air weignt perday.(Ref.3). This lea age rate, used in the evaluation of offsite doses resulting from accidents, is defined in 10 CFR 50, Appendix J (Rev.1), as L,: the maxinun allowable containment leakage rate at the calculated peak containment internal pressure (P ) resulting from the l

s WR. TheallowableleakageraterepresentedbyL

[dbyh /wf 4#Af limiting forms tf $basisfortheacceptancecriteriaimposedonall, containment leakage rate testing. L, is assumed to be b

(continued)

WOG STS B 3.6-2 Rev 1, 04/07/95 hc4Mr ,

9 l-

l McGuire & Catawba Improved TS Review Conssents

/ ITS Section 3.6, Containment Systems D 3 TE.2 Containment Air Locks '

3.6.2-1 DOC A.5 (ITS 1.0)

CTS 1.7 See Comment Number 3.6.1-3. Comment: Revise the CTS Markup of ITS 3.6.2 to include a markup of CTS 1.7. Provide additional discussions and justifications for the Administrative changes.

DEC Response:

See response to Comment 3.6.1-3.

REVISED RESPONSE: ,

The CTS page is added, however, the intent of the definintion is captured by the LCO itself, therefore, the change is a format change, DOC A1, as discussed in the comment resolution meeting June 17,1998.

O l

/^\

V mc4_cr_3.6- 3.6-5 July 2, 1998 9

Freeddg, g,(,,s l DEFINITIONS /

'CONTAINMLni mii'G^iTV

, O I

() 1.7 CONTAINMENT INTEGRITY shall exist when:

l a. All penetrations required to be closed during accidept conditions are either:

et. ned n oC. g is yg*l 1) Ca able of being closed by an OPERABLE containment automatic is lation valve system, or operator action during periods when containment isolation valves may be opened under administrative l controls pursuant to Specification 4.6.1.1.a; or

2) Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions.

, All equipment hatches are closed and sealed, L.c4 ff,Q c. Each air lock is in compliance with the requirements of Specification 3.6.1.3,

h. The containment leakage rates are within the limits of Specification 3.6.1.2, and

e. Thesealingmechanismassociatedwitheachpenetration(e.g., welds, bellows,or0-rings)isOPERABLE.

CONTRbbLEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor ]

coolant pump seals.

, CORE ALTERATION

/;, ITI (,6 1.9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

l CORE OPERATING LINITS REPORT 1.10 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document l that provides core operating limits for the current operating reload cycle.

These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.9. Unit operation within these operating limits is. addressed in individual specifications.

l McGUIRE - UNIT 1 1-2 Amendment No. 166 i.

O en .c c

L

\ sp e:CM;,3.s.t DEFINITIONS <

]

IEONTAINMENTINTEGRITY 1.7 CONTAINMENT INTESRITY shall exist when:

a. All penetrations required to be closed during accidept conditions are either: .

L Secuy 1) Capable of being closed by an OPERABLE containment automatic p, g M*g isolation valve system, or operator action during periods when containment isolation valves may be opened under administrative controls pursuant to Specification 4.6.1.1.a; or

2) Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions.

, , All equipment hatches are closed and canlad ~

C c.ph,4Nj gU 3gNr4 . Each air lock is in compliance with the requirements of Specification 3.6.1.3, T The containment leakage rates are within the limits of Specification 3.6.1.2, and

e. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPER&BLE.

CONYELLEDLEAKnut %

18 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor olant pump seals.

RE ALTERATION

' D .9 CORE ALTERATION shall be the movement or manipulation of any component fee wh ithin the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude complettlen of g" M I.* movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT

, 1.10 TheCOREOPERATINGLINITSREPORT(COLR)istheunit-specificdocument that provides core operating limits for the current operating reload cycle.

These cycle-specific core operating Ilmits shall be determined for each reload cycle in accordance with Specification 6.9.1.9. Unit operation within these W ratino limits ic =ddressed in individual specifications.

McGUIRE - UNIT 2 1-2 Amendment No. 148 l

k w_ -

McGuire & Catawba Improved TS Review Comments ITS Section 3.6, Containment Systems 3.6.2-2 DOC A.9 DOC LA.3 ~

DOC LA.4 JFD 6 JFD Bases 6 CTS 3.6.1.3.b CTS 4.6.1.3a, b, and d.

ITS SR 3.6.2.1 ITS B3.6.2 Bases l

See Comment Number 3.6.1-2. Comment: See Comment Number 3.6.1-2.

DEC Response: .

DOC LA.3 and LA.4 have been revised to indicate that the CTS 4.6.1.3.a. 4.6.1.3.b, and 4.6.1.3.d detail have been relocated to the Bases. The frequency of testing is already established by 10 CFR 50, Appendix J, Option A and is not necessary to be repeated in the surveillance. The Bases provide an appropriate place to locate the details of the a:alleak testing which are in addition to the requirements of the CFR. ITS SR 3.6.2.1 and associated Beses are revised to conform to the STS. DOC A.9 has been revised to delete references to OV the leak rata testing program and to Option B. ITS 5.5.2 and associated CTS markup have been revised to delete the air lock acceptance criteria from the Containment Leakage Rate Testing Program.

REVISED RESPONSE:

During the comment resolution meeting with NRC June 17,1998, a number of consistency issues were identified which are discussed below.

1. Note 2 for SR 3.6.2.1 is revised to conform to the markup for TSTF-52.

2. . (McGuire Only) Brackets are removed from STS markup SR 3.6.2.1 item a acceptance criteria. '

I

3. (McGuire Only) Bases Page B 3.6-27 of the STS markup, last paragraph for )

SR 3.6.2.1 should reference SR 3.6.1.2 in lieu of 3.6.1.1, consistent with the actual {

SR. i i

July 2,1998 inc4_cr_3.6 3.6-6 9

I

j. Containment Air Locks l-3.6.2

, SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR -3.6.2.1 -------------------NOTES-------------------

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

l

2. Results shall be evaluated against -----NOTE------

acceptance criteria applicable to SR 3.0.2.is not SR 3.6.1.2. applicable Perform required air lock leakage rate In accordance testing in accordance with 10 CFR 50, with 10 CFR 50, Appendix J. Option A, as modified by Appendix J, approved exemptions. Option A, as modified by l The acceptance criteria for air lock approved testing are: exemptions

a. Overall air lock leakage rate is

.g s 0.05 L, when tested at 2 P.,

c Q .m

b. For each door, leakage rate is < 0.01 l ~L when tested at 2 14.8 psig.

l

,,^

SR- 3.6.2.2 -

Perform a pressure test on each inflatable 6 months l air lock door seal and verify door' seal l 1eakage is < 15 sccm.

i' M, SR 3.6.2.3 Verify only one door in the air lock can be 18 months j; opened at a time.-

i i-p

\

.l

.D.. .

'McGuire Unit 1 3.6-7 Supplement 6 l r

a e .- .

\

r Containment' Air Locks 3.6.2 O SURVEILLANCE REQUIREMENTS L) SURVEILLANCE FREQUENCY SR 3.6.2.1 -------------------NOTES-------------------

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

2. Results shall be evaluated against -----NOTE------

acceptance criteria applicable to SR 3.0.2 is not SR 3.6.1.2. applicable Perform required air lock leakage rate In accordance testing in accordance with 10 CFR 50, with 10 CFR 50..

Appendix J, Option A, as modified by Appendix J, approved exemptions. Option A, as modified by The acceptance criteria for air lock approved testing are: exemptions

a. Overall air lock leakage rate is s 0.05.L when tested at 2 P.,

b. For each door, leakage rate is < 0.01 L, when tested at 2 14.8 psig.

'SR 3.6.2.2 Perform a pressure test on each inflatable 6 months air lock door seal and verify door seal

-leakage is < 15 secm.-

SR 3.6.2.3 Verify only one door in the air lock can be 18 months opened at a time.

O' g McGuire Unit 2 3.6-7 Supplement 6 l

-_______-_m. _ - _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _

Specification 3.6.A Insert 5 s A3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.1 ---____--_-----_--NOTES--------------------

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

' ------NOTE----

2. Results shal1 be evaluated against SR 3.0.2 is not acceptance criteria applicable to 3R applicable.

3.6.1.2. .___..._______.

Perfonn required air lock leakage rate In accordance A testing 4 accordance with10 CFR 50, with 10 CFR 50, V Appendix J. Option A, as modified by Appendix J, approved exemptions. Option A, as modified by The acceptance criteria for air lock approved testing are: exemptions

a. Overall air lock leakage rate is < 0.05 La when tested at E Pa-

b. For each door, leakage rate is <

l 0.01 La when tested at t 14.8 psig.

I l

l l

f O.

k~ McGuire 1 Page 3 of f

'9 L_____ _ _ _ _ _ _ . . _ _ _

(:;

i 1

rw. Specification 3.6. 2 Insert 5 A ,cj ;

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.1 ------------------NOTES--------------------

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

, ------NOTE-----

2. Results shall be evaluated against SR 3.0.2 is not acceptance criteria applicable to SR applicable.

3.6.1.2. ---------------

Perform required air lock leakage rate In accorda.nce

(-)

(f testing in accordance with10 CFR 50, Appendix J, Option A, as modified by with 10 CFR 50, Appendix J, approved exemptions. Option A, as modified by The acceptance criteria for air lock approved testing are: exemptions

a. Overall air lock leakage rate is 5 0.05 La when tested at t Pa-

b. For each door, leakage rate is <

0.01 La when tested at t 14.8 psig.

p McGuire M Page 5 off-~

f s

_A____ _ _ _ _ _ . _ _ _ _ _ . - _ . - - . . . - - _ _

o. '

l Containment Air Locks rnw % e. w w ,*=rie Ice Condenser. ana <ImT1 3.6.2 @ f SURVEILLANCE lEQUIIBDils

, SURVEILLANCE FREQUENCY 1'

- SR 3.6.2.1 -------- MTES---- ---

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

feskagetest.ormance of the overall y air lock _ ,

i> 2. Results shall be evaluated'ac'e1' acceptance crite __ SR 3.6.1 '1 i RWi= ?=T Pe orm required air lock leakaoe rate j'

r--

f i t---- I 7

testing in accordance with uunnuq $R .0.2 is' t "9y 4-a s -~_J mi^y approved icable (xer,1ti s. Ep - - - - -

The a aptance crite a for air loc . In Ec =a. -e 0 h testeY at P.

. ch r l k e is D qw9 g.f '"'

ea ag r ff;;-

1 53.6.2[f-- - - - ---NOTE - ' --- l l Only requi exit th to be perf the contai upon ent air loc or' QMj

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

)( lo Verify only one door in the air lock can be f1K"iEphk opened at a time.

^

WOG STS- 3.6-7 Rev 1. 04/07/95 L

,k

6 Containment Air Locks a6s=anaric. woatmosener,c. ice t,on==ar w nuIn 8 3.6.2 h

l BASES (continued) ;

l s < u s -

8 SINWEILLAICE REQUIREIDITS

' SR 3.6.2.1 N h', h .

C-f I

Maintaining containmen air locks OPERABLE requires coupliance with the kage rate test _ requirements of h n . M. 9 W/ig lardpHana _- - _ _ w m --m .m m - u wa ov moor -

yx ou,f This SR reflects the lenka rate testino W itA*"'P" requirements with regard to air lock 1 (Type B leakage tests). The acceptance criteria were est lished durino initial air lock and containment OPERABILITY testing. The periodic testing requirements verify that the air lock leakage does not exceed the allowed fraction of the overall O(,- . _ containment leakege rate. The Frequency is required by ,

5 . Y Y Y 1 5 5 $ TensTonTdo$'noh)ff M (s.ssWA1) -a-

'The SR has been modified by two Notes. Note 1 states that an inoperable air lock door does not invalidate the previous successful perfoneance of the overall air lock leakage test. {.

This is considered reasonable since either air lock door is I capable of providing a fission product barrier in the event of a DBA. Note 2 has been added to this SR requiring the O M

.g

,5 y results to be evaluated against the acceptance criteria SR 3.6.1/k -lhis ensures that air lock leakage is proper y -

accounted for in determining the MdFM containment leakage 4

g @*- c - . a tve. s a p l SR 3.6.2.2 @h h 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 i withstand the maximum ted post accident containment oressure, closure of eit r door will support containment r OPERABILITY. Thus, the door interlock feature sgperts h containment OPERABILITY while the air lock is bei personnel transit in and out of the containment. riodic Sused ' for d@

1 testing of this interlock demonstrates that the interlock /

will function as designed and that simultaneous opening of

% mild-A the inner and outer doors will not inadvertently occur. Due to the curely mechanical nature of_this interlock, and given met . snet the interloca mechanian is enry> challenged when the w.H ee4=

e * =y(*=i'""

enntainment air lock door ist153Em this test is only lo a,,,. ) _

N I T_ Nk .NEM "M _

(continued)

I . .

101STS. I _ B 3.6 *>7 1. 04/07/95 f eMe,y if a Rs. %e,r. is ,-My,,A, a s , Rev O . % keg.} . is te,a , % e a .

,o lXJ wade,. A c 2.I,Ls o.4 apply Joe f o pf 4-a p 4c.NeI for for o IO5t I8 i L u?Mt.aE'.4*3a ,u >I o,4.,a.,J a,. Fa no,ory u . M#

oi. 1 - .r e ,4 J44c/t 6 e (*(** b

e res.

g,. ped + r'efl : %i

/

l l McGuire & Catawba Improved TS Review Comments ITS Section 3.6, Containment Systems 3.6.3 Containment isolation Valves .

3.6.3-1 DOC A.5 (ITS 1.0)

CTS 1.7 See Comment 3.6.1-3. Comment: Revise the CTS markup of ITS 3.6.3 to include a markup of CTS 1.7. Provide additional discussions and justifications for the Administrative changes.

. DEC Response:

See response to comment 3.6.1-3.

i REVISED RESPONSE: ,

The CTS page is added, however, the intent of the definintion !s captured by SRs 3.6.3.3, j 3.6.3.4, and 3.6.3.7, therefore, the change is a format only change, DOC A1, as discussed in !

the comment resolution meeting June 17,1998.

l l

!O

(,) mc4_cr_3.6 3.6-11 July 2, 1998 L__--__-_-_--_---__---___--_--_ - - _ - - H

DEFINITIONS 3*

y

CONTAINMENT INTEGRITY O

h 1.7 CONTAINMENT INTEGRITY shall exist when

a. All penetrations required to be closed during accident conditions are either: *

(cybJ 5't M 1. b l g/ 1) Capable of being closed by an OPERABLE containment automatic

\ / mL.,JW isolation valve system, for operator acuon auring perious wnen -

Q- N t L= *R I. containment isoiau on valves may be opened under administrative C.,L,dt19 controls pursuant to Specification 4.6.1.1.a; or SR.14.1.3

+ 2) Closed by manual valves, blind flanges, or deactivated automatic fit 3.( .s.9 valves secured in their closed positions.

F

. b. All equipment hatches are closeu ano sealea,

c. Each air lock is in compliance with the requirements of k Specification 3.6.1.3 i

d. The containment leakage rates are within the limits of Specification.

3.6.1.2, and

e. Thesealingmechanismassociatedwitheachpenetration(e.g., welds, bellows, or 0-rings) is OPERABLE.

Set. =L4 \

(gg CONTR0Lt.ED LEAKAGE

, 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

I i-Q CORE ALTERATION

.V l 1.9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT 1.10 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle.

l These cycle-specific core operating limits shall be determined for each reload l cycle in accordance with Specification 6.9.1.9. Unit operation within these Qeratinglimitsisaddressedinindividualspecifications.

l I

I, McGUIRE - UNIT 1 1-2 Amendment No. 166

\

iV 13d13 f

DEFINITIONS SPCCdi=EJo3le.3 CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

a. All penetrations required to be closed during accident conditions are

, either: .

y s Ec,3h 1) Ca able of being closed b an nornanir containment auta= tic cvLJb is lation valve system rator action during periods when' i

_ containment isolation valves may be opened under administrative w%

a,L,JL ;3 Q controls pursuant to Specification 4.6.1.1.a; or j

SL 3.t 43 y 2) Closed by manual valves, blind flanges, or deactivated automatic j g 79,y,q valves secured in their closed positions.

b. All equipment hatches are closed and sealed, j

c. Each air lock is in compliance with the requirements of i Specification 3.6.1.3,

d. The containment leakage rates are within the limits of Specification 3.6.1.2, and

e. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPER&BLE.

E' *k L 05, t.s CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

( COltE ALTERATION 1.9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT 1.10 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle.

These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.9. Unit operation within these ~

l operating limits is addressed in individual specifications.

L ~

l McGUIRE - UNIT 2 1-2 Amendment No. 148 A

Qy UO 0

McGuire & Catawba Improved TS Review Comments ITS Section 3.6. Containment Systems 3.6.3 DOC A.1 JFD 1 -

I~ JFD Bases 1

' CTS 4.6.1.2.e (Catawba) -

CTS 4.6.1.2.f (McGuire)

ITS SR 3.6.3.7 and Associated Bases LITS B3.6.3 Bases SR 3.6.3.5 CTS 4.6.1.2.f specifies that the combined bypass leakage rate shall be determined "during .

each Type A test" This material has not been retained in ITS SR 3.6.3.7. The proposed change has been categorized as a reformatting, renumbering, or rewording type change. As a result of the discussion in Comment Number 3.6.1-2, this phrase cannot be deleted, but .

. must be included in some form in the frequency for ITS SR 3.6.3.7. In addition, the Bases for ITS SR 3.6.3.5 and SR 3.6.3.7 need to be modified in accordance with Comment Number 3.6.1-2. Comment: See Comment Number 3.6.1-2.

DEC Response:

The phrase in question only applies to penetrations which are not individually testable, in these cases, the. CTS allows these penetrations to be tested with soap bubbles during Type A' O

testing while the containment is pressurized. The Type A tests.are performed by 10 CFR 50 J Appendix J, Option B.iTherefore, the frequency of the Type A tests will not be the same as

for Types B and C testing which is performed for the other penetrations. The frequency has -

been modified to require penetratbns not individually testable to be tested during Type A -

testing consistent with the CTS requirements.

REVISED RESPONSE:

During the comment resolution meeting with NRC' June 17,1998, a number of consistency issues were identified which are discussed below.

3 ;1. :The STS markup for Bases SR 3.6.3.8 includes inserted text which only addresses the

~ frequency for penetrations which are individually testable. The existing STS text is :

acceptable for both testable and non-testable penetrations, therefore, the text is deleted and the STS is adopted.

' 2.

~

The STS Bases for SR 3.6.3.6 and SR 3.6.3.8 make references to 10 CFR 50

, Appendix J. Option A is added to clarify which part of Appendix J is applicable to these Type B and C tests.

b

~'

~ mc4_cr_3.6- 3.6-12 July 2, 1998

. l\

. \l ,

__ __ _ a

i

' Containment Isolation Valves B 3.6.3 I

\

[J BASES (continued)

SURVE'ILLANCE SR 3.6.3.4 (continued)

REQUIREMENTS Frequency of " prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days" is appropriate since these containment isolation valves are operated under l

administrative controls and the probability of their misalignment is low. The SR specifies that containment isolation valves that are open under administrative controls are not required to meet the SR during the time they are open. This SR does nat apply to valves that are locked, sealed, or otherwise secured in the closed position, since i these were verified to be the correct position upon locking, sealing, or securing.

This Note allows valves and blind flanges located in high radiation areas to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, 3, and 4, for ALARA reasons. Therefore, the probability of misalignment of these containment isolation valves, once they have been verified to be in their proper position, is small.

SR 3.6.3.5 Verifying that the isolation time of each automatic powei-operated containment isolation valve is within limits is required to demonstrate OPERABILITY. The isolation time test ensures the valve will isolate in a time period less than or equal to that assumed in the safety analyses. The isolation time is specified in the UFSAR and Frequency of this SR are in accordance with the Inservice Testing Program.

SR 3.6.3.6 For containment purge valves with resilient seals, additional leakage rate testing beyond the test requirements of 10 CFR 50, Appendix J, Option A, is required to ensure l OPERABILITY. The measured leakage rate for contair.nent purge lower compartment and instrument room valves must be s 0.05 L, when pressurized to P,. The measured leakage rate for containment purge upper compartment valves must be s 0.01 L, when pressurized to P,. Operating experience has

/"T '(continued)

V McGuire Unit l' 8 3.6-27 Supplement 6 l

Containment Isolation Valves B 3.6.3

( BASES (continued)

SURVEILLANCE SR 3.6.3.8 I REQUIREMENTS (continued) .This SR ensures that the combined leakage rate of all reactor building bypass leakage paths is less than or equal to the specified leakage rate. This provides assurance that the assumptions in the safety analysis are met. The leakage rate of each bypass leakage pcth is assumed to be the-maximum pathway leakage (leakage through 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. 1n 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 valves in the penetration are closed, the actual leakage rate is the lesser leakage rate of the two valves. This method of quantifying maximum pathway leakage is only to be used for. this SR (i.e., Appendix J maximum pathway leakage limits are to be quantified in accordance with Appendix J). Penetrations which are not individually testable shall be determined to have no detectable leakage when tested with soap bubbles while the containment is pressurized during SR 3.6.1.1 Type A testing. The Frequency A is required by 10 CFR 50, Appendix J. Option A, as modified Q by approved exemptions (and therefore, the Frequency extensions of SR 3.0.2 may not be applied), since the testing is an Appendix J. Type B or C test. This SR simply imposes additional acceptance criteria.

Bypass leakage is considered part of L,.

REFERENCES 1. UFSAR, Section 15.

2. 10 CFR 50.36. Technical Specifications, (c)(2)(ii).

3. UFSAR, Sec.tton 6.2.

4. Generic Issue B-24.

l-

5. Standard Review Plan 6.2.4.

l McGuire Unit 1 B 3.6-29 Supplement 6 l

\

l Containment Isolation Valves l B 3.6.3 BASES (continued)

! SURVEILLANCE SR 3.6.3.4 (continued)

REQUIREMENTS Frequency of " prior to entering MODE 4. from MODE 5 if not performed within the previous 92 days" is appropriate since these containment isolation valves are operated under administrative controls and the probability of their misalignment is low. The SR specifies that containment isolation valves that are open under administrative controls are not required to meet the SR during the time they. are open. This SR does not apply to-valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be the correct position upon locking, sealing, or securing.

This Note allows valves and blind flanges located in high radiation areas to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, 3, and 4 for ALARA reasons. Therefore, the probability of misalignment of these containment isolation valves, once they have been verified to be in their proper position, is small.

IR 3.6.3.5 Verifying that the isolation time of each automatic power operated containment isolation valve is within limits is required to demonstrate OPERABILITY. The isolation time test ensures the valve will isolate in a' time period less than or equal to that assumed in the safety analyses. The isolation time is specified in the UFSAR and Frequency of this SR are in accordance with the Inservice Testing Program.

i SR 3.6.3.6 i For containment purge valves with resilient seals, additional leakage rate testing beyond the test requirements of 10 CFR 50, Appendix J, Option A, is required to ensure l OPERABILITY.- The measured leakage rate for containment ,

purge lower compartment and instrument room valves must be l s 0.05 L, when pressurized to P,. The measured leakage rate for containment purge upper compartment valves must be s 0.01 L, when pressurized to P,. Operating experience has (continued)

D'w)

McGuire Unit 2 a 3.6-27 Supplement 6 l

Containment Isolation Valves B 3.6.3 BASES (continued)

SURVEILLANCE SR 3.6.3.8 I REQUIREMENTS (continued) This SR ensures that the combined leakage rate of all reactor building bypass leakage paths is less than or equal to the specified leakage rate. This provides assurance that the assumptions in the safety analysis are met. The leakage rate of each bypass leakage path is assumed to be the maximum pathway leakage (leakage through 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 valves in the penetration are closed, the actual leakage rate is the lesser leakage rate of the two valves. This method of quantifying maximum pathway leakage is only to be used for this SR (i.e., Appendix J maximum pathway leakage limits are to be quantified in accordance with Appendix J). Penetrations which are not individually testable shall be determined to have no detectable leakage when tested with soap bubbles while the containment is pressurized during SR 3.6.1.1 Type A testing. The Frequency is required by 10 CFR 50, Appendix J, Option A, as modified O by approved exemptions (and therefore, the Frequency extensions of SR 3.0.2 may not te applied), since the testing is an Appendix J, Type B or C test. This SR simply imposes additional acceptance criteria.

Bypass leakage is considered part of L,.

REFERENCES 1. UFSAR, Section 15.

2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

3. UFSAR, Section 6.2.

4. Generic Issue B-24.

5. Standard Review Plan 6.2.4.

O McGuire Unit 2 e 3.6-29 Supplement 6

, l

ContainmentIsolationValvesJAtmMpher71c O. GiToaunosnner,c. icsEnndentar^ndnely@

B 3.6.3 p ; %,R m.

BASES NS. c'ne p.ce u dt(akca tw eW *d nJehe-

.,3w a-v.iw % 4 5 0.07 L.

.se., ,,essunu J 4. Pa . th.mtsi m J SURVEILLANCE SR 3.6.3. t..tw r.k R. c M.A F@ urre e REQUIREMENTS tr ap w e8hs o,og ,ae.

(continued) For containment purge valves with resilient seals #*" Mite. J additional leakage rate testing beyond the test requirements J A ._ of 10 CFR 50. Annendix J, is required to ensure OPERABILITY.(

' Operating experience nas demonstrated that this type of seal has the potential to degrade in a shorter time period than do other seal types. Based on this observation and the importance of maintaining this penetration leak tight (due (e to the direct path between containment and the environment),

a Frecuency of 184 days was establishedKs part of t N be c resputionofGenericissuef-20,'ContainmentLea e (to heal Deterioration" (Ref. 3). f -

[dditio ly, this W @ t ha nartorme(within 92 gays atty

-open' the valve d The 92 day Frequency was cnosen recognizing that cycling the valve could introduce *

.- additional seal degradation (beyond that occurring to a nt ab + pury. valve that has not been opened). Thus, decreasing the 9,. cm ,,4,...A vales interval (from 184 days) is a prudent measure after a valve ey be us I dtr.e.

a".. I has been opened. Y--(ru com.cm.c,. ~r7r3e. Ioiver cesparhent pu% m.,% , In adcM* 0h : **'** * **d 'n s + r nse e+ roo m va.1v es re rnain f, the_,sfV cWj reguenef R 3"6'3"

=be.,(edorns d% arma l of" anct this this SR must \ ,,(U % P"~"M Mg W darj4 fee ev*r9 98. dat' av,i' ^

y oncsc. valves.

Automatic containment isolation valves close on a -

or*^'aj the Ja 'V 8- containment isolation signal to prevent leakage.of radioactive material from containment followin a DBA. This SR ensures that each automatic containment iso ation valve will actuate to its isolation position on a containment k 6 M 8*h e isolation signal.1 This surveillance is not required for Gy*h I"E i valves that are locked, sealed, or otherwise secured in the dre fhW A> r uired sition under administrative controls. The h f 6* g ,g g

/4 7"Fp

/ f 1 mont Frequency is based on the need to erform this Surveillance under the conditdces that apply uring a plant t--

outage and the potential for an u.i lanned transient if the Surveillance were performed with t e reactor at power.

Operating experience has shown that these cowponents usually pass this Surveillance when performed at the *181 month Frequency. Therefore, the frequency was concluded to be acceptable from a reliability standpoint.

(continued)

'00 STS-B 3.6-42 Rev 1, 04/07/95 M<.65e- .

f'T ContainmentIsolationValvesJAtmosyEgl h

_. ___ __ _ _..

B 3.6.3 BASES

$ ~

l SURVEILLANCE SR 3.6.3. d (continued) -

REQUIREMENTS' maxiasn pathway leakage (leakage through 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 6e the actual pathway leakage through the isolation device. If r both isolation valves in the penetration are closed, the k actual leakage rate is the lesser leakage rate of the two valves. This method of quantifying maxima pathway leakage <

is only to be.used for this SR (i.e.. Appendix J maximum /

pathway leakage limits are to be quantified in accordance ( Qi haddi.n J A '" ** witis Appendix J).YThe Frequencytis reauired by 10 CFR 50 / g .

1.td, L :. $ 4+" kt mooenaix 4 as modified by approved ex **

J gh .. d k t,..< e. r tnereTore, the Frequency extensions 3.0.2of may S tions not be( h ,g . . \. n gay,i, L.Ly Jen applied), since the testing is an Appendix J Type test. ' l'.-

Uj g,) J.it. nq,4.W'S SR simply imposes additional acceptance criteria. 369

'*"""' [

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i L

L "P 2f "*" 4'; 3-4 sv nass leakaoe is considered nart of L,. IRev.$cefrS

Olote:Arnless soecifissTlf exempted)T) l O . sg.s.t,.O -

I G '

Id u ri^3

__1. @fSAR, Section '115L l

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f A @SAR Section 16.2L.

l Cr C ric Issue B-20, 'Corgainment Leakage Due/f o Seay b

\ erioration." e

4. Generic Issue B-24.

45. standard WevacuJ Fla n 4.2. C M m cAq .So.3co, 7dthna*u./ Speu$a Sted1S WE j

l I

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l E STS B 3.6-44 Rev 1, 04/07/95

^ jticfmsthb l

McGuire & Catawba Improved TS Review Comments

.ITS Section 3.6, Containment Systems i

3.6.3-5 DOC A.18 -

JFD 11 JFD Bases 5 CTS 4.6.1,1.a -

CTS 4.6.3.2 ITS SR 3.6.3.2 and Associated Bases .

ITS SR 3.6.3.3 and Associated Bases

. ITS SR 3.6.3.6 and Associated Bases CTS M6.3.2 is revised to include a clarification that valves which are locked, sealed, or

- otherwise secured in their required safety position are not required to be tested. CTS

. 4.6.1.1.a requires verifying that all penetration not capable of being closed by an OPERABLE automatic isolation valve and required to be closed during accident conditions are closed.

CTS 4.6.1.1.a becomes ITS SR 3.6.3.2 and ITS SR 3.6.3.3. Both of these ITS SRs and their -

associated Bases are revised per TSTF 45 to include a clarification that valves which are locked, sealed or otherwise securing in their safety position are not required to be verified closed. The CTS does not contain this provision. Therefore CTS 4.6.1.1.a needs to be revised to include this provision and the. change designated DOC A.18. See Comment

' Number 3.6.3-6. Comment: Revise the CTS markup of CTS 4.6.1.1.a to include this

. Administrative change and revise DOC A.18 to discuss and justify this change to CTS

~4.6.1.1.a.

DEC Response:

The CTS markup and DOC A.18 are revised.

REVISED RESPONSE:

4 :

The CTS markup does not show the proposed change in the text for 4.6.1.1.a for penetrations located outside containment. DOC L34 is added to justify this change.

i l

l~ mc4_cr_3.6 3.6-15 July 2, 1998

'i'. _a_ L L____ _ _ _ -

1.

. spec & at G g.G.3

.- ~- .

4.6 CONTAIIDIDtT SYSTDtS 3/4.5.1 PRIMARY coltTAlleIDfT ,

CONTAfletDrT INTEGRITY

[ ree Ndeb LfilfTillE COISTTI0ff FOR OPDtATI0ff i

l h /T5 3.(r.f / 3.6.1.1 primary collTAtl01DIT IllTEGit!TT shall be maintafaed.

APPLICABILITY: H0 DES 1, 2, 3, and 4. "*

f

$13: '

. itithout primary CollTAlleIDIT IllTEElt!TT, rettere CollTAllelDIT IllTEGit!TY within I hour erwithis SIRmlonel be latheatfollowing least HOT STAlW8T 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . withis the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and la COL j g aggtAner neOureDiDirs -

/N5eff // .6.1.1 primary CollTAlleIDtf IllTEGit!TT shall be demonstrated SR 3.6 3, @_ At least once per 31 day by verifyfog that all .penetrattens* not */

i capable of beleg closed l

pt4

-- Se 3.68.h valves or operator acti OPDtASLE costalament autenatic isolation during periods aheal.patainment isolation valves are open under administrative controlft9and required to me '

.JfwQ eAQg win,ilJS9,,%

A flanges, or deactivated automatic v lclosed during accident l veks% ~

a ves secured in their positions;

[ n.leJ,,,4.W'aa<t takS,sW A*fjf b. By verifying Specification 3.6.1.38 and that rach costalament air lock is la compliance with

' c. After each closlag of each Sed Mrd'kjd except the containment air tration subject to Type 8 testtag, 4# g,0*j test, by leak rate testing the se,al with gas at14.8 Pkspsig, if opened and followies a verifylag that 'when the seasured leakage rate for.,hese t seals is added to the leakage rates detersized pursuant te S 4.6.1.2d. for all other Type 8 and C penetrations, specification leakage rate is less than 0.60 L,. the cashined h.h S# f.6,3.M *Except valves, blind flanges, and deactivated automatic k ivalves dich are J c.YeJ M

S83 63* located faside the contaissent and the annulus mdisW1ocked, sealed or #'*d

~

etherwise secured in the closed position. These penetrationsys%nll be 1 1 vertfled closed during each COLD $151f00ls1 except that such vertftcation need not h performed more eften than esce per 92 days.

- The fell valves may \

.M 4 trati rois IIC-14 en an infMaittent basi [under adel -

-42,ifE-13.)lE'-23,VX-34,ylc40,fw.11 13 ,

C++A ene-time change is granted to have the contalement purge supply and/

enha==t isolatica valves for the upper and louer coopertment open la Modes U8 3 and 4 followleg the steam generator replacement estage. The cumulative g,.17.5 3*4, f time for barlag the valves open in Modes 3 and 4 is llatted to fourteen -

(14) days. All other previsions of this specification apply with the esception of these containment purge valves open la Modes 3 and 4. Each we will be sealed closed prior to initial entry late Mode 2.

l McGuitE - IIIIIT 1 3/4 6-1 Amendrent fle.174 ff 6'E G i

I t-

- - - - - - - ____ ~~ ____ --

, , . == QCt!/ht A S.k* 3 3/a.6 C0tTAfpMENT SYSTDfS 314.6.1 PRIMARY CONTAflMDrT CONTAllMDff INTEGRITY See Marktf LINIMC MIM

O' ERA

4 17534,/ 3.5.1.1 Primary CONTAllMENT INTEGRITY shall be maintained.

Appl 1CABftTTY: MODES 1, 2, 3, and 4. *"

EllE! i itithout primary CONTAllMENT INTEGRITY, restore CONTA!! within MENT INTEGIIITY 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least NOT STAleBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in SIRJTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . ~

' SURVEfttANCE REOUTREMENTS L,7 Q.6.1.1

~ primary CONTAtlMENT INTEGRITY shall be demonstrated: [

fjg;gpf // '

50* 6*a At least once per 31 days by verifying that all penetrasfons* not L.')V valves or operator action during periods whertsontainme c.g 3,u,sQ) valves are open under administrative contro15d and reqvfred t f* 4Nagh*dladM t closed during accident conditions are closed by vaim, s o be

'iind- *g MaA & olvu N4 - gflanges, sy verif or deactivated automatic valves secured in their positions;

~,

e. N kea,:= 4 a.,.,.

obau hyg, [b. .spectffegon3.6.1.3;andthat each containment air lock is in compitance with

~

c. 1

.See. Norg Y\N After each closing of each penetration ~ subject to Type 8 testing,

' test, except the containment air locks, if opened following a Type A or 5 g j75 y"4,f O by leak rate testing the seal with gas at p

~

added to the leakage rates determined pursuant to sverifying t

Qeakage rate is less than 0.60 L,.,4.6.1.2d. for all other Type 8 a -

N >t, '

ce te O'4'gg 2363,.3@

located faside the containmen,t and the annulus iDe d4NF scaled or "seced, and are lock verffled closed during each COLD SHLITDOWN except tha need not be performed more often than once per 92 days.

I L..T The following valves may be openes on an intermittent. basis under admints-T

)FW-4.trative control IIC-141, NC-142 WD13,11623, VI.34, YX-40, FW-11, FW-13, i 2,se *"A one-time change is granted to have the containment p 3 and 4 following the steam generator replacement outage.

,fef./ j3 ".Q/ time for having the valves open in Modes 3 and The 4 iscumulative limited to fourteen (14) exc days. All other provisions of this specification apply with the keptian of those containment purge valves open in Modes 3 and 4.

Each) will be sealed closed prior to fattial entry into Mode 2.

MesutRE - UNIT 2

/

3/4 6-1 Amendment me.156 C) ex 6 ' !*

e

Discussien of Changes Ssctien 3.6 - Containment Systems TECHNICAL CHANGES - LESS RESTRICTIVE F only eliminates a' duplication of requirements and the CFR provides sufficient regulatory control over this activity. This change is consistent with NUREG-1431.

L.34 CTS SR 4.6.1.1.a requires each penetration located outside containment to-be verified closed by a valve, blind flange, or deactivated automatic valve once per 31 days. ITS SR 3.6.3.3

. excludes penetrottons that are locked, sealed or otherwise secured in position from this verification. This change is

, considered acceptable since any penetration that is secured in l- the correct position is meeting the safety functton. These penetrations would not be inadvertently misposittoned since they l= ore secured in the closed position. This change is consistent i^ with CTS exceptions provided for other valve alignments, e.g.

! ECCS, which are locked, sealed, or otherwise secured in position. This change is also consistent with NUREG-1431.

L.35 CTS 4.6.3.1 Surveillance Requirement requires an operability verification for each containment isolation volve prior to retu u tng the valve to service after maintenance, repair or

( replacement work on the volve or its associated actuator, or l control or power circuit. These details describing when post maintenance testing is required to be performed is redundant to L' controls in the QA Topical Report and is not necessary for inclusion within the TS. Changes to the QA ^ Topical Report are

, evaluated under the requirements of 10 CFR 50.54. Any change, using this criteria, will ensure proper review and confon,mnce to the QA Program requirements. This change is consistent with

~ NUREG-1431.

L.36 The- CTS 4.6.4.2.b.1; requirement for a channel calibration of the hydrogen recombiner is deleted. The hydrogen recombiner instrumentation does not relate directly to the system OPERABILITY and is not necessary for inclusion within the TS. Control of the ovatiability of, and necessary compensatory activities if not . j available, for indication instruments, monitoring instruments, and '

alarms are addressed by plant operational procedures and policies which are controlled by the plant procedure control program. In addition,- the system functional test required by ITS SR 3.6.7.1 will ensure that'necessary controls will functton properly. This

change.is consistent with NUREG-1431.

O l 1

McGuire Units 1-and 2 Pa'g e' L - 1243 Supplement 63 l !

N3 Signifiest H:zards C:nsideratien S:cticn 3.6 - Containment Systems 1 jm LESS RESTRICTIVE CHANGE L.34 The McGuire Nuclear Station is converting to the Improved Technical l Specifications (ITS) as outlined in NUREG-1431, ' Standard Technical Specifications, Westinghouse Plants." The proposed change involves l making the current Technical Specifications (CTS) less restrictive.

Below is the description of this less restrictive change and the No Significant Hazards Consideration for conversion to NUREG-1431. j l

' I CTS 59 4.6.1.1.a requires each penetration located outside !

containment to be verified closed by a valve, blind flange, or deactivated automatic volve once per 31 days. ITS SR 3.6.3.3 i excludes penetrations that are locked, sealed or otherwise secured

'in position from this verification. This change is considered acceptable since any penetration that is secured in the correct position is meeting the safety function. These penetrations would not be inadvertently mispositioned since they are secured in the closed position. This change is consistent with CTS exceptions provided for other volve alignments, e.g. ECCS, which are locked, sealed, or otherwise secured in position. This change is also consistent with NUREG-1431.

O In accordance with the criteria set forth in 10 CFR 50.92, the McGuire Nuclear Station has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following is provided in support of this conclusion.

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

The proposed change exempts penetration status verification for penetrations which are locked, sealed, or otherwise secured in the closed position. The penetrations which are secured are passive devices and are not considered initiators of any analyzed event.

Therefore, the probability of an accident previously evaluated is not significantly increased. The proposed change does not reduce the performance requirements or acceptance criteria for the containment since the penetrations are secured in the assumed closed position, therefore, the consequences of analyzed events

~

are not affected.

O v

McGuire Units 1 and 2 P ge 6B M of 73 M Supplement 63 l

N) Significant Hazards C:nsid:ratien i S:cticn 3.6 - Containment Systems t

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

The proposed change does not permit operation in a new or different mode, or permit the installation of a new or different type of equipment. The proposed change exempts containment penetration verification for penetrations secured in the closed position. The containment remains capable of performing the design safety functions. Therefore, the proposed change does not create the possibility of a new or different kind of accident from those previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The proposed change continues to require the same performance and acceptance criteria assumed within the safety analysis for the containment penetrations. The allowance to exclude penetrations which are locked, sealed, or otherwise secured in the closed position from position verification does not affect these assumptions. Therefore, this change does not involve a (m) significant reduce a margin of safety.

'V lMcGuireUnits1and2 P ge 69M of 73M Supplement 63

(

McGuire & Catawba ~ Improved TS Review Comments ITS Section 3.6, Containment Systems D) 3.6.3-11 JFD Bases 4 STS B3.6.3 Bases - RA A.1 and A.2 l

STS B3.6.3 Bases - RA E.1, E.2 and E.3 STS B3.6.3 Bases - SR 3.6.3.3 ITS B3.6.3 Bases - RA A.1 and A.2 ITS B3.6.3 Bases - RA E.1, E.2 and E.3 ITS B3.6.3 Bases - SR 3.6.3.2 STS B3.6.3 Bases for RA A.1 and A.2, RA E.1, E.2 and E.3 and SR 3.6.3.3 describe how the verification of the correct position of the containment isolation valves is to be performed. This l verification is performed via a system walkdown. The ITS modifies these Bases sections to state that the verification is "through administrative controls such as a system walkdown or computer status indication." The proposed change is unacceptable and changes the intent of the verification. The first part of the change (" administrative controls such es") would not limit

~

the means of verification to only a system walkdown or computer status indication; a paper l verification would be allowed by this change. This is unacceptable to the staff. In addition, no discussion or justification is provided to describe what is meant by computer status indication and why it is equivalent to a system walkdown; it is assumed that the computer status .

indication is merely a status listing (paper verification) which would be unacceptable.

Comment: Delete both parts of this change, or provide additional discussion and justification to show that the computer status indication is the equivalent of a system walkdown.

DEC Response:

The proposed Bases have been revised to eliminate a reference to administrative controls.

The CTS requires that valve position be verified but does not specify the methods for performing the verification. The operator aid computer (OAC) and control. board indications provide the operator with valve position status for a number of valves. These indications are considered equivalent to a system walkdown for the purposes of verifying system alignment.

Therefore, the proposed ITS Bases which allows computer status indication for determining valve position is acceptable and consistent with the current requirements. Valves which do not have control room status indication are verified through system walkdowns.

REVISED RESPONSE:

The OAC monitors various process parameters and equipment status throughout the plant, including valve position, through the use of installed process instrumentation. Operators can detemine actual valve position for a number of valves via the indications available in the control room from the OAC. These indications are considered equivalent to a system

. walkdown for the purposes of verifying system alignment.

L

! v mc4_c r_3.6 3.6-21 July 2, 1998 9

L.

t j ..

McGuire & Catawba Improved TS Review Comments ITS Section 3.6, Containment Systems I

3.6.5-2 JFD Bases 3 (Catawba)

JFD Bases 4 (McGuire)

STS B3.6.5 Bases - SR 3.6.5.1 and SR 3.6.5.2 ITS B3.6.5 Bases - SR 3.6.5.1 and SR 3.6.5.2 STS B3.6.5 Bases - SR 3.6.5.1 and SR 3.6.5.2 states that "The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency for these SRs is considered acceptable based on observing slow rates of temperature increase within containment as a result of environmental heat sources (due to the large volume of the containment)." This material has not been adopted in the Bases for ITS B3.6.5. Justification JFD Bases 3 (Catawba) and JFD Bases 4 (McGuire) do not specifically address this proposed j

. deletion, and give different justifications for the same change. Comment: Provide additional discussion and justification to show why this STS statement is not applicable to McGuire and Catawba and revise the ITS Bases markup accordingly.

DEC Response:

l The surveillance frequency is adequately justified by other statements in the STS Bases. This statement is not considered accurate since ice condenser containments are relatively small in comparison to other containment types. This appears to be a carryover statement from the other containment designs included in the STS which do have large free volumes. JFD 13 has been added to justify this deletion.

REVISED RESPONSE:

Although the ice condenser containment is small, in discussions with the NRC staff at the June 17,1998 comment resolution meeting, the staff considers the statement appropriate.

l The Bases are revised to conform to the STS and JFD 13 is deleted.

i i

bV 3.6-26 July 2, 1998 mc4_cr_3.6

Containment Air Temperature B 3.6.5 BASES SURVEILLANCE SR 3.6.5.1 and SR 3.6.5.2 (continued) $

REQUIREMENTS containment ventilation unit. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency of these SRs is considered acceptable based on observed slow rates of temperature increase within containment as a result of environmental heat sources (due to the large volume of containment). Furthermore, the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is considered adequate in view of other indications available in the control room, . including alarms, to alert the operator to an abnormal containment temperature condition.

REFERENCES 1. UFSAR, Section 6.2.

2. 10 CFR 50.'49,

3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

O l

O l .McGuire Unit 1 S 3.6-38 Supplement 6 L

Containment Air Temp 3rature B 3.6.5 O BASES'-

SURVEILLANCE SR 3.6.5.1 and SR 3.6.5.2 (continued) I REQUIREMENTS containment ventilation unit. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency of these SRs is considered acceptable based on observed slow rates of temperature increase within containment as a result of environmental heat sources (due to the large volume of containment). Furthermore, the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is considered adequate in view of other indications available in the control room, including. alarms, to alert the operator to an abnormal containment temperature condition.

REFERENCES 1. UFSAR, Section 6.2.

2. 10 CFR 50.'49.

3. 10 CFR 50.36. Technical Specifications, (c)(2)(ii).

(continued)

.O j McGuire Unit 2 .B 3.6-38 Supplement 6

1 r

Justification f:r Deviatiens S:ctien 3.6 - R2 fueling Operaticns l BASES 1

/

12. The changes are consistent with generic change TSTF-30 to NUREG-1431 provided to NRC by the industry owners groups,

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lMcGuireUnits1and2 I 32 Supplement 63

,,~

f a v-Containment Air Temperature OceAnmimarru -

B 3.6.5@ Q BASES ACTIONS B.1 and B.2 (continued) does not apply. To achieve this status, the plant nust be brought to at least K DE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to K)DE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based en cperating experience, to reach the required plant conditions from full power conditions in an-orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6 M.1 and SR 3.6.M.2 @

REQUIREMENTS Verifdng that containment average air temperat ire is within the LW limits ensures that containment operation remains 43,jg , within the limits assuned for the containment analyses. In

- order to determine the containment average air temperature, f b/f4 5/=Laa5 M ## . weighted averagelis calculated using measurements taken at locations within tF.e containment selected to provide a

/

\ representative sample of the overall containment atmosphere _. / /

\ggg/

/7

. } T6e c., nou rrequenc f these is wuswe o acceps based on oserved s'y rates of emperature nerease w' hin le' l ( '

t as a sult of en ronmental t source (due f

t contai ito t large vol of cont t). F hermore Ig

'N 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is consloered adequate in view o other indications available in the control room. including alarms, to alert the operator to an abnormal containment temperature condition.

REFERENCES 1. @SAR. Section f6.?A. h

2. 10 CFR 50.49.

4 I

(to cat 60.56, McAngg }cfaltoms,(c)(2)GL&

4 ,

p B 3.6-59 Rev 1. 04/07/95 He (qsu**O n

f

_~

McGuire & Catawba Improved TS Review Comments

-( ITS Sectico 3.6, Containment Systems e

3.6.6-3 JFD Bases 4 ITS B3.6.6 Bases BACKGROUND The discussion for ITS B3.6.6 Bases-BACKGROUND states in the insert on Page B3.6-87 that the ph of the sump solution is raised to at least 8.0 within one hour of the onset of the LOCA. The insert material also states that the chemical mixing tank and the charging pumps are used to accomplish this but it does not describe how this accomplished. Comment:

Revise ITS B3.6.6 Bases BACKGROUND insert discussion to address this issue.

DEC Response:

This discussion has been misinterpreted. The resultant pH of the sump solution is based on the mixing of the RCS fluids, ECCS injection fluid, and the melted ice which are combined in the sump as a result of a postulated pipe break and subsequent blowdown. The ITS Bases insert material indicates that it is also possible to adjust this resultant pH using the charging pumps and mixing tanks. The procedural details of how this is accomplished are not relevant to this Bases discussion for the containment spray system and are a level of detail not consistent with the STS Bases for the generic material which was replaced with plant specific material.

I C~

O REVISED RESPONSE:

-The Bases have been revised to delete the sentence about use of the mixing tank and charging pump and a sentence has been added describing the resultant pH.

t.

(_

( mc4icr_3.6 3.6-29 July 2, 1998

Containment Spray System

, B 3.6.6 1

O BASES V

BACKGROUND The Containment Spray System and RHR System provide a spray (continued) of cold or subcooled borated water into the upper containment volume to limit the containment pressure and temperature during a DBA. The RWST solution temperature is q an important factor in determining the heat removal capability of the Containment Spray System during the injection phase. In the recirculation mode of operation, heat is removed from the containment sump water by the Containment Spray System and RHR heat exchangers. Each train of the Containment Spray System, supplemented by a train of RHR spray, provides adequate spray coverage to meet the system design requirements for containment heat removal.

For the hypothetical double-ended rupture of a Reactor Coolant System. pipe, the pH of the sump solution (and, consequently, the spray solution) is raised to at least 8.0 within one hour of the onset of the LOCA. The resultant pH of the sump solution is based on the mixing of the RCS fluids, ECCS injection fluid, and the melted ice which are l combined in the sump. The alkaline pH of the containment sump water minimizes the evolution of iodine and the occurrence of chloride and caustic stress corrosion on fm mechanical systems and components exposed to the fluid.

i The Containment Spray System is actuated either i automatically by a containment pressure high-high signal or I manually. An automatic actuation opens the containment l spray pump discharge valves, starts the two containment spray pumps, and begins the injection phase. A manual actuation of the Containment Spray System requires the operator to actuate two separate train related switches on the main control board to begin the same sequence of two train actuation. The injection phase continues until an RWST level Low-Low alarm is received. The Low-Low alarm for the RWST signals the operator to manually align the system to the recirculation mode. The Containment Spray System in the recirculation mode maintains an equilibrium temperature between the containment atmosphere and the recirculated sump water. Operation of the Containment Spray System in the recirculation mode is controlled by the operator in accordance with the emergency operation procedures.

l i.

I O (continued) b l - McGuire Unit 1 .B 3.6-40 Supplement 6

Containment Spray System B 3.6.6

~ BASES BACKGROUND TheContainmentSpraySystem.andRHRSystemproIideaspray (continued)- of cold or subcooled borated water into the upper containment volume to limit the containment pressure and temperature during a DBA. The RWST solution temperature. is an important factor in determining the heat removal capability of the Containment Spray System during the injection phase. In the recirculation mode of operation, heat-is removed from the containment sump. water by the Containment Spray System and RHR heat exchangers. Each train of the Containment Spray System, supplemented by a train of RHR spray, provides adequate spray coverage to meet the system design requirements for containment heat removal.

For the hypothetical double-ended rupture of a Reactor Coolant System. pipe, the pH of the sump solution (and, consequently, the spray solution) is raised to at least 8.0 within one hour of the onset of the LOCA. The resultant pH of the sump solution is based on the mixing of the RCS fluids, ECCS injection fluid, and the melted ice which are combined in the sump. The alkaline pH of the containment sump water minimizes the evolution of iodine and the occurrence of. chloride and caustic stress corrosion on mechanical , systems and components exposed' to the fluid.

O' The Containment Spray System is actuated either automatically by a containment pressure high-high signal or manually. An automatic actuation opens the containment spray pump discharge valves, starts the two containment spray pumps, and begins the injection phase. A manual actuation of the Containment Spray System requires the operator to actuate two separate train related switches on the main control board to begin the same sequence of two train actuation. The injection phase continues untti an RWST level Low-Low alarm is received. The Low-Low alarm for the RWST signals the operator to manually align the system to the recirculation mode. The Containment Spray System in the recirculation mode maintains an equilibrium temperature between the containment atmosphere and the recirculated sump water. Operation of the Containment Spray System in the recirculation mode is controlled by the operator in accordance with the emergency operation procedures.

(continued) l: McGuire Unit 2 8 3.6-40 Supplement 6

_L_____-_---__----.._ _ --- - - - - - '- ' " - ~ ~ ~ " ~ ~ ~ '

O INSERT @

For the hypothetical double-ended rupture of a Reactor Coolant System pipe, the pH of the sump solution (and, consequently, the spray solution) is raised I to at least 8.0 within one hour of the onset of the LOCA. The resultant pH of I the sump solution is based on the mixing of the RCS fluids, ECCS injection 1 fluid,'and the melted ice which are combined in the sump.

i O

insert ea,e , ,.d.,,

O ,c ,,,,

I

McGuire & Catawba Improved TS Review Comments ITS Section 3.6, Containment Systems

l. '3.6.6 JFD Bases 4 STS B3.6.6 Bases - SR 3.G.6.1 ITS B3.6.6 Bases - SR 3.6.6.1 STS/ITS SR 3.6.6.1 verifies that each containment spray manual, power operated, and

' automatic valve in the flow path that is not locked,' sealed, or otherwise secured in position is -

in the correct position. STS B3.6.6C Bases-SR 3.6.6C.1 states that this verification is to be

. accomplished by a system walkdown for those valves outside containment. ITS B3.6.6 Bases

- SR 3.6.6.1 modifies the STS wording to allow Administrative controls to be used for the verification. The Administrative controls would include a system walkdown, but would also allow a paper verification, a computer status indication or any other administrative means of l- verifying valve position. This change significantly modifies the intent of the SR. See Comment Number 3.6.3-11. Comment: See Comment Number 3.6.3-11.

DEC Response:

1 .

L The proposed Bases have been revised to eliminate a reference to administrative controls.

l The CTS requires that valve position be verified but does not specify the methods for performing the verification. The operator aid computer (OAC) and control board indications provide the operator with valve position status for a number of valves. These indications are O considered equivalent to a system walkdown for the purposes of verifying system alignment.

L.k) Therefore, the proposed ITS Bases which allows computer status indication for determining valve position is acceptable and consistent with the current requirements. Valves which do not have control room status indication are verified through system walkdowns.

_1 REVISED RESPONSE:

The OAC monitors various process parameters and equipment status throughout the plant, including valve position, through the use of installed process instrumentation. Operators can detemine actual valve position for a number of valves via the indications available in the control room from the OAC. These indications are considered equivalent to a system walkdown for the purposes of verifying system alignment.

l- . . . .

3.6-30 July 2, 1998 mc4_cr_3.6 9

9

l McGuire & Catawba Improved TS Review Comments ITS Section 3.6 Containment Systems

?

3.6.12-3 DOC LA.14 CTS 4.6.5.1.b.2 (McGuire)

CTS 4.6.5.1.c (Catawba)

. ITS SR 3.6.12.2, SR 3.6.12.3 and Associated Bases (McGuire)

ITS SR 3.6.12.4, SR 3.6.12.5 and Associated Bases (Catawba)

CTS 4.6.5.1.'b.2 (McGuire) and CTS 4.6.5.1.c (Catawba) require verifying that the minimum

. : average ice weight of a representative sample of ice baskets shall not be less than 1081 (McGuire)/1273 (Catawba) pounds per basket at a 95% level of confidence. The CTS markup indicates that "at a 95% level of confidence" was not retained in corresponding ITS SR - ,

e ' 3.6.12.2.a and SR 3.6.12.3 for McGuire and SR 3.6.12.4.a for Catawba but instead was !

relocated to the Bases. However, "at a 95% level of confidence " was in fact retained in corresponding ITS SR 3.6.12.2a, and SR 3.6.12.3 for MCGuire and SR 3.6.12.4.a and SR

' 3.6.12.5 for Catawba. The CTS markup is in error. Comment: Revise the submittal to correct the CTS markup.

DEC Response:

Duke Energy disagrees that the CTS markup is incorrect. CTS 4.6.5.1.b.2 (McGuire) and

~

. CTS 4.6.5.1.c (Catawba) do specifically require the weight of an additional 20 baskets be determined at a 95% confidence if the representative 6 baskets contain less than the required amount.- These CTS surveillance ara retained in ITS SR 3.6.12.4.a. The ITS surveillance only discusses representative baskets and relocates the discussion of 6 baskets and 20 bast:sts and the required confidence level to the Bases for ITS SR 3.6.12.4. A specific statement for the 95% confidence level is not specified in ITS 3.6.12.4.a. consistent with the STS. ITS SR 3.6.12.4.b does specify the 95% confidence level, but this is from the discussion of total ice weight (not the representative sample) which comes from a different paragraph in the CTS surveillance. 4 1

REVISED RESPONSE:

The CTS markup for McGuire unit 1 incorrectly indicates that the 95% confidence level statement for the ice weight sample surveillance (CTS 4.6.5.1.b.2, ITS 3.6.12.3) is moved to the Bases per DOC LA14. This discussion is retained in ITS 3.6.12.3 and the CTS markup is l corrected to show this information retained.

O mc4_cr_3.6 3.6-44 July 2,1998 i

6pec4 kes.&dv1 5.fo. I2.

~

j

< sf CMrAlleeDrT _SYS7D45 SuevtfLLAnct REINIRDEITS (Castlemed) flbasketeachf tel hous I, 2, 4, 6, 8. and g (or ren

' ) the same row of adjacent bay if a basket free a de gnated few cannot be feed for weighing) within each boy If any ,g basket is f to contata less than 1001 po e ts of ce. A r.r..: W ive le of It addittenal baskets f the same bar shall be i . The stalm e everage wel S h of Ice from the to addi one beshets and the discrepent be shall not

@ 1ess _ tant mandsMet at a 9% leve of ceaffdence.)

l' '

The ice condenser shall also be subdiv{ded late 3 roups of R 3,6./2. besbets, as follows: Group 1 - Bays 1 through 8, reap 2 and Grene 3 Says 17 through 24 The snei-i r

toys evere,e num 9 through 16.lyht of the sample baskets free aestas ice we news 1, 2, 4. L 4. and e in each ereup shall set than i

1981 pounds h abet me 4 954 levew m - " - l S$2T {

The minism total ice condenser ice weight at a SH level of 4 M. 3.4,/1. 2.} confidence shall be calculated using all _ ice basket weights _

L detersteed ductee this wishlee orgram tes shah met be less j .ft, 5 to . IL. L * ** ** * * F*****5 *** \

54. 3 6.fl. 4 @ Verifying. by a- visual lespecties _' ar nr

' maer = - - .$

mae .-- ._ -- -

,__r#; L^_:zaL*** ; -: ratia!*.*r**

7.-wlation of frvst er f cemi .i..w .

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Wit %IGdV6-a thickness ef less thee E_--or semel to ..

0.38 Igch.yl 7 - W .. : ^ :.;;- frost er co m phf,. q sus j ,

Esce with a thidaevs of prester than er equal to e.as inch, a representative sample of 20 additteest flew passages free the gc[g i S r1P M S C 'c 6 same bay shall be elseelly inspected. If these additlanal flom Ipassages are food acceptable. the surveillance program may ( nde++se P 'I I

proceed coesidertog the single deficiency as untene and accept i

Ale. Iterv than ene restricted flew pasg per bay is cidence of sheeruel deeradottee of the 1ce condenser.

I gp_ $ 4. /2. 6, @ At least_ence per 40 meeths or nrn== ame'rtsastly lespecttEE"Yll>

I J-- 2 i tme Ice baskets from each -

sces= 7,,.

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

detrimental structural meer, cracks. corrosies, or, ether damage. _

n u - ; n . .. se reises at r e- - .m . m i, .---i MM r,. .c Ie m at . Wi!T 1 3/46-21 Ammatumt No. 166 C S. 0 5 9

y'

.. .McGuire & Catawba Improved TS Review Comments p ITS Section 3.6, Containment Systems.

., V d

3.6.13 DOC M.4 CTS 3.6.5.3 ACTION b, ITS 3.6.13 ACTION D CTS 3.6.5.3 ACTION b requires that with one or more ice condenser doors inoperable (not

. capable of automatic opening) and not restored to OPERABLE status, that the unit be in i MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , MODE 4 within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within the following 30'

_ , . hours. ITS 3.6.13 only requires the unit to be MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The requirement to be in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of the commencement of a shutdown has been deleted. The justification (DOC M.4) used states that the deletion is a More Restrictive change. This is incorrect. - The change is Less Restrictive, since the ITS does not specify a time limit in which to reach MODE 4. Comment: Revise the CTS markup and provide additional discussion and justification for this Less Restrictive change.

DEC Response: .

The overall change is 'more restrictive since the CTS allows a total of 42 hours4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months to reach MODE 5 (out of the mode of applicability) and the iTS only allows 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months to reach MODE 5.

It is not the practice of the ITS conversion to dissect every phrase, but rather to capture the substance of the change. in this case, the actions are directing a unit shutdown and the O with respect to a unit shutdown. desired MODE is MODE 5. Therefore, the ITS is more REVISED RESPONSE:

DOC M4 is revised to provide this additional discussion.

b d mc4_cr,,3.6' ,

3.6-50. . July 2, 1998 j i

l' t._i__________1____J___!_ - _ _ - - - _ - _ - - - - - - ~

l Discussicn cf Chang;s j 5:cticn 3.6 - C ntainment Systems j ) TECHNICAL CHANGES - MORE RESTRICTIVE M.1 CTS 3.6.1.3 Action a.1 is revised to establish a one h'our time limit for verifying the OPERABLE air lock door is closed, when the other door is inoperable. CTS requirements do not specify a time.

One hour is a reasonable period of time for this verification because the status of the air lock doors is indicated in the control room. The change, retained in ITS 3.6.2, is more restrictive and consistent with NUREG-1431.

M.2 CTS 3.6.1.3 Action b provides actions to address any air lock inoperability other than an inoperable door. This includes having two air lock doors in the same air lock inoperable at the same time. These actions are revised to add an immediate action to evaluate the overall containment leakage rate requirement of LC0 3.6.1, " Containment," and establishes a one hour time limit to verify one air lock door closed in the affected air lock. The CTS does not specify a time limit for closing one door. These additional actions are more restrictive and consistent with NUREG-1431.

M.3 An action is added to CTS 3.6.3 and 3.6.1.9 to verify that (Vl isolated penetration flow paths remain closed. These are penetrations that have been isolated due to an inoperable isolation valve. The actions require isolation valves be verified every 31 days and prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days. In addition, valves with resilient seals are required to be tested every 92 days when the purge valves are closed to isolate an inoperable isolation valve.

These are additional requirements and are considered more restrictive and consistent with NUREG-1431.

M.4 CTS 3.6.5.3 oction b requires a shutdown to Mode 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , Mode 4 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and Mode 5 by 42 hours4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months . The pccific ecmpletien time to recch MODE 4 in CTS 3.5.5.3 cctica b Sc; been deleted. The CTS pecified cn intermediate step in the shutdown

~

action to reach Mode 4 which wc; isnot necessary and not consistent with similar actions in this specification and in other specifications nor with the requirements specified in LC0 3.0.3.

The actions in ITS 3.6.13 require the unit be placed in MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 5 in the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Although the deletion of the specific Mode 4 shutdown time could be considered

('M less restrictive, Tthe overall 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months reduction in total

() completion time to reach Mode 5 is explicityly more restrictive.

lMcGuireUnits1and2 Page M - la Supplement 63

I-Discussien cf Chingss SIcticn 3.6 - Containment Systems l p.

1V TECHNICAL CHANGES - MORE RESTRICTIVE l

l Therefore, the overall change is more restrictive, but is consistent with similar shutdown actions throughout the CTS and with NUREG-1431.

l M.5 Not used.

.M.6 Not used.

M.7 An additional surveillance is proposed for CTS 3.6.1.7 to verify each door in each access opening to the reactor, building is closed, except when the access opening is being used for normal transit entry and exit. This additional surveillance is acceptable because the integrity of the reactor building is important to the dose calculations following a DBA and it is required by CTS 1.27.a. This change is considered more restrictive because it requires additional surveillance beyond that which are presently performed for the reactor building integrity. This change, retained in ITS 3.6.16, is consistent with NUREG-1431.

O M.8 CTS 3.6.4.3 Actions provide actions for inoperable hydrogen O igniters. If these actions 'can not be satisfied, CTS LCO 3.0.3 must be entered which requires action within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or a shutdown to MODE 3 in the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . ITS 3.6.9 requires that if these same actions cannot be met the plant mut.c be placed in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . This change is slightly more restrictive since the CTS would provide 1 additional hour. This change is acceptable because it places the plant in a MODE where the specification is not applicable. This change is consistent with NUREG-1431.

M.9 CTS 4.6.5.3.2.a requires verifying that the intermediate deck doors are free of frost accumulation. ITS SR 3.6.13.2 requires verifying that the doors are not impaired by ice, frost, or debris. This requirement is more restrictive than the current stated requirements, however, it is consistent with existing operating practices and with NUREG-1431.

O J

McGuire Units 1 and 2 Page M - 23 Supplement 63l

McGuire & Catawba Improved TS Iteview Comments ITS Section 3.6 Containment Systems 3.6.13-6 JFD Bases 11 STS B3.6.16 Bases - RA B.1 and B.2 and RA C.1

- ITS B3.6.16 Bases - RA B.1 and B.2 and RA C.1.

The last sentence in STS B3.6.16 Bases - RA B.1 and B.2 states the following: "If this j verification is not made Required Actions D.1 and D.2 not Required Action C.1 must be 1 taken."_ in addition, the last sentence in STS B3.6.16 Bases RA C.1 states the following:

" Condition C is entered from Condition B only when the Completion Time of Required Action B.2 is not met or when the ice bed temperature has not been verified at the required frequency." Both of these statements have been deleted from ITS B3.6.16 Bases - RA B.1 and B.2 and RA C.1_ respectively. The justification for this deletion (JFD Bases 11) states that the Bases discussions are not consistent with the specification nor with the rules of Completion Times as defined !n NUREG Section 1.3. The staff believes that the two statements are correct and neea M remain. The staff's interpretation of the statements is that if the ice bed temperature is not surveilled in accordance with the frequency limitation j specified in ITS SR 3.0.2 due to forgetfulness or inattention to ACTION requirements, rather l than inability to perform surveillance, a shutdown is required, rather than allowing an additional 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months _to restore the ice condenser door to OPERABLE status. In addition, the ,

' staff considers the change to be generic and beyond the . scope of review for this conversion.

Comment: Delete this generic change.

i d DEC Response:

The staff's interpretation is neither consistent with the current technical specification l requirements, STS 3.6.16 as written, STS LCO 3.0.2, nor with the rules of Completion Times (

as described in STS 1.3. The proposed interpretation is also inconsistent with the actions for '

the ice bed temperature LCO 3.6.12. The stated action in STS 3.6.16, required action B.1,is to verify ice bed temperature is within limits on a periodic frequency of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and required action B.2 requires the inoperable doors be restored in 14 days. If at any time during the 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months , the temperature is not within limits, then required action C.1 becomes applicable and requires the ice condenser be restored in 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . This is exactly the same action as ;

required by STS 3.6.15, Required Action A.1. This action is also applicable (and intentionally identical) because with the temperature !imit not met, the LCO is not met and the actions of STS 3.6.15 become applicable. In both STS 3.6.16 and 3.6.15, a completion time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months

- is allowed with the ice bed temperature not within limits if the temperture limit is not checked,

the STS Bases provide conflicting statements which also do not agree with the actions as 1 written.' The Bases for required action B.1 and B.2 state that if the verification is not made,

. condition D applies. The Bases for required action C.1 states that it applies when the temperature has not been verified at the required frequency. Condition D, however, clearly indicates in the Specification that it is only applicable to Conditions A and C.

- CTS 3.6.5.3 allows operation for up to 14 days provided the ice bed temperature is monitored every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and is within limits; otherwise, the doors must be restored in 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The

.mc4_cr_3.6 3.6-52 July 2. 1998 S

_ _ _ _ _ . _ _ _ _ __________,.___m_._ _ ___ _. _ . _ _ _ . _ -_ ._ ___ .__. _ _ . _ ~ . . _ _

l McGuire & Catawba Improved TS Review Comments j ITS Section 3.6, Contaimeent Systems STS Bases cannot establish new rules for the usage of completion times which,are not consistent with those already established by STS 1.3, nor can the Bases direct actions which are in direct conflict with the actions of the LCO as written. The STS Bases is incorrect on both counts and is a more restrictive change on the current license. Duke Energy does not l accept this more restrictive charge for inclusion in the ITS, nor is this change from the STS

. Bases considered generic because it maintains the current licensing basis.

REVISED RESPONSE:

I JFD 11 is revised to include the above discussion.

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l mc4fr_3.6 3.6-53 July 2, 1998 l .

Justificati:n far Deviaticns Sectirn 3.6 - Refueling Operations A

BASES

10. The changes are consistent with generic change TSTF-17 to NURfG-1431 l provided to NRC by the industry owners groups, except that an 18 month frequency, consistent <ith the current refuel cycle is proposed.

11. The discussion in the NUREG 3.6.16 Bases for Actions B.1, B.2, C.1 and C.2 are incomplete and not consistent with the Specification nor with the rules of completion times as defined in NUREG Section 1.3. The discussion indicates that Condition C is only entered for limited cases of not meeting required Action B.1. This is not represented in the specifications and is also not consistent with the first sentence in the Bases discussion for C.1 and C.2. The Bases for ITS 3.6.13 provides the missing detail and removes the incorrect discussions. The NUREG Bases are also inconsistent with action' sfor the ice bed temperature LCO 3.6.12.

The stated action in STS 3.6.16, required action B.1, is to verify ice bed temperature is within limits on a periodic frequency of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and required action B.2 requires the inoperable doors be restored in 14 days.

If at any time during the 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months , the temperature is not within limits, then required action C.1 becomes applicable and requires the ice condenser p be restored in 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . This is exactly the some action as required by STS 3.6.15, Required Action A.1. This action is also applicable (and t} intentionally identical) because with the temperature limit not met, the LCO is not met and the actions of STS 3.6.15 become applicable. In both STS 3.6.16 and 3.6.15, a completion time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is allowed with the ice bed temperature not within limits. If-the temperture limit is not checked, the STS Bases provide conflicting statements which also do not egree with the actions as written. The Bases for required action B.1 and B.2 state that if the verification is not made, condition 0 applies. The

. Bases fc required action C.1 states that it applies when the temperature has not been verified at the required frequency. Condition 3, however, clearly indicates in the Specification that it is only applicable to Conditions A and C.

CTS 3.6.5.3 allows operation for up to 14 days provided the ice bed temperature is monitored every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and is within limits; otherwise, the doors must be restored in 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months . The STS Bases cannot establish new rules for the usage of completion times which are not consistent with those already established by STS 1.3, nor can the Bases direct actions which are in direct conflict with the actions of the LCO as written. The

- STS Bases is incorrect on both counts and is a more restrictive change on the current license. Duke Energy does not accept this more restrictive change for inclusion in the ITS, nor is this change from the STS Bases Q' Q . considered generic because it maintains the current licensing basis.

McGuire Units 1 and 2 '

23 Supplement 63 l

'McGuire & Catawba Improved T5 Review Comments

'ITS Section 3.6, Containment Systems 3'6.14-2.

. DOC A.29 t

CTS 3.6.5.5 ACTION (McGuire)

; CTS 3.6.5.5 ACTION a (Catawba)

ITS 3.6.1.4 Condition A Note

, ; CTS 3.6.5.5 ACTION (McGuire) and ACTION a (Catawba) provide the requirements in the E

event that a personnel access door or equipment hatch is inoperable or open except for l . personnel transit entry. The CTS markup indicates that this is ITS 3.6.14 ACTION A. The Note for ITS 3.6.14 Condition A states that " separate Condition entry is allowed for each

personnel access door or equipment hatch."_ The proposed change has been categorized as an Administrative change. '.This is incorrect. The wording of the CTS 3.6.5.5 ACTIONS do not indicate that a separate condition entry is allowed in the CTS, as would be allowed in the containment isolation valve CTS. In this case, if more than one access door or hatch is INOPERABLE, CTS 3.0.3 is entered. Thus, the proposed change is Less Restrictive.

Comment: Revise the sut,ndital and provide the appropriate discussion and justification for this Less Restrictive change.

-DEC Response:

Duke Energy disagrees with this interpretation. The " separate condition entry" notes were O added during development of the STS to be consistent with those cases where both the

Lindustry and the staff agreed that separate condition entry was appropriate and permitted l- under the current technical specifications due to an agreement that the STS would be condition based rather than component based (refer to Westinghouse Owners Group comments on the draft STS provided to NRC July 1991). In this case, the actions are g

currently interpreted to allow the action time to apply to each access door or hatch individually based on the language used, i.e., "with a... inoperable" instead of "with one... inoperable." This is more clearly seen in the Catawba LCO which uses different language in the_two actions.

l . That is, the action for doors and hatches states "with a ... door or equipment hatch (other than one pressurizer enclosure hatch) inoperable..." and the action for the pressurizer enclosure hatch states "with one..." It is clear that the requirements are distinctly different in the existing actions. ' Duke Energy does not believe that the proposed ITS is less restrictive than the g current licensing basis with respect to the application of this action.

1'

_ REVISED _ RESPONSE:

DOC A29 is deleted and DOC L38 is added to justify the change as less rest? lve.

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July 2, 1998 mc4_cr_3.6 3.6-57 P

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drsonnei .uess doorde.C3v,'f;"'W!.sdA.TC' j

,4p e h.tchBeperme status er , encgept for 4.so Md personnel transit entry, restors the door er hatch te erraamsr e its closed position gas espiscoern vithis 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months sur n in et seast nui 5 misnin sus was e neurs ans in COLD $NWTDOlst within the following 30 4

sunwrsttascr arnutesersis ,

kgMcD64 A 3,(,,,/% / 6 Thecontalauset's persannel access deers and agul batches hetmeen the c44evi upper and louer compartments sha 1 he determined closed _

visual inspection prior tegneressium sue agector casiant system Ig

-Q d2OO'F mas ..~ . . - r.._ a transts m., amen sm --ser 6eesens ayssee .

s 4 J.E*.tv. ,5 I, Is abose ISB*F. 4 dA 3. de-(C/ E GI23 The personnel access deers arW tgulpm.ust hatches between the castainment's upper and lamer comW L Aall he determined OPERABLE by visually inspecting the seats and se. sing urfaces of these penetrotsens and verifytag me detreematal afsalipments, crects er defects in the seeling serfaces, or apperset deterioretten of the seal esterials

a. . prior to final closure of the penetratten each time it has been l

pened. and

b. At least once ser steet le yearsNor enterials. penetrations containing seeg

,) g ranricates from rest _

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personnel transit entry. resters the door er hitch to OplEAeLE states # to tic ** a ses c.,

- - sma. I h r; 6 . es i . wi staessy withte the nest 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to ceLa SNU1DRel withle the (ellenlag M Acynat/ C,, heers.

SiggrtsIaarf eEEUIRBENIS Ibe persensel access deers and egel f4. ff,,Na [ @inment's

_ conta esser and louer espertmentsbatches she between determined the closed by e 88NM f4006 4 !

tignal leapecties prior to0 u-. .e === y w.i- er .u L essee) [doen M0066 r -

. r..- u-is ri - -- wei ses ,i. Ael gggg, fY'g N The persesnel access deers and equi hatches hetuose the

-- costeinment's and louer campertments she deterstand ertansLE by visselly tospect the seals and seelles surfaces of these penetrations and verffytog as I sisell , cracts er defects le the seeling serfeens, er apparent esters les of the seal esterials

a. Prior to final closure of the penetrstica each tem It has been 8pened,and

h. At 1eest enee per to nee c.erpenetrettees costelelag see s a

y enricem rres restilent astertels.

fesutet . IIIIT 2 3/4 6-25 Ansedmont no. les ay / ,/ v

Discussien of Ching2s i

Section 3.6 - Containment Systems

(

( ADMINISTRATIVE CHANGES A.29 Not used.A Note 1 added t0 CTS 3.5.S.S which pc: :it;' cparate Ccndition entry for cc S perscnnel ;;ce;; door er equip cnt hatch.

Thi ch:nge in ne ::::ry to provid explicit instruction; for pr per- pplication Of the ACT!0NS for Technical Specif! : tion

plf;nce. In conjunction with ITS 1.3, "C:=pleti^- Time ," thi N0te pr;;fde directi;n ecn i; tent with the intent of the exi ting ACTIONS f;r in;per:ble pers nnel :::::: d;;r er equipment hatch and i; ther fere, 00n;fdcred administrative. Thi: change, retained in ITS 3.5.11, i n;f tent with N' REC J 1131.

A.30 CTS 3.6.5.5 Action requires inoperable or open access doors or equipment hatches to be restored to OPERABLE status "or" closed (as applicable) depending on whether a hatch or door was inoperable. ITS 3.6.14 requires them to be restored to OPERABLE status "and" closed. The change removes the ambiguity and provides clarification that the door or hatch must be closed to maintain divider barrier integrity and does not specify any new requirement. This wording change does not alter the existing requirements established in the surveillance requirements and is (y therefore, considered administrative. This change is consistent with NUREG-1431.

A.31 CTS 3.6.5.9 Action requires an inoperable divider barrier seal be restored to OPERABLE status prior to increasing the Reactor Coolant System temperature above 200 'F but does not provide

]

actions if the seal is inoperable after the RCS is above 200 'F.

This requires entry into LC0 3.0.3 which provides I hour to correct the problem. ITS 3.6.14 requires an inoperable seal be restored to OPERABLE status within I hour. The ITS requirement is the same as entering the CTS LC0 3.0.3, therefore, this change is considered administrative. This change is consistent with NUREG-1431.

A.32 CTS 3.6.5.7 Action requires an inoperable floor drain be restored to OPERABLE status prior to increasing the Reactor Coolant System temperature above 200 *F but does not provide actions if the seal is discovered inoperable after the RCS is above 200 'F.

Therefore, the CTS would require entry into LC0 3.0.3. LC0 3.0.3 would require actions within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or a shutdowr.. ITS 3.6.15

[ requires an inoperable drain be restored to OPERABLE status within O

V 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or a shutdown. Since the ITS requirement is the same as i

. i McGuire Units 1 and 2 Page A - 810 Supplement 63l l

)

Discussien cf Chang 2s Section 3.6 - Containment Systems p

N TECHNICAL CHANGES - LESS RESTRICTIVE L.37 The CTS 4.6.1.7 requirement to report abnormal degradation

. discovered during reactor butIding survetIlances is deleted. The change is acceptable since inspections and reporting of degradation ore required by 10 CFR 50.55a and are not necessary for inclusion within the TS. Requirements associated with regulations are not necessary to be duplicated in the TS and are implemented directly. = These changes are consistent with NUREG-1431.

L.38 CTS 3.6.5.5 actions requires an inoperable personnel access door or equipment hatch be restored operable or closed in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . CTS 3.0.3 would be applicable if more than one hatch or both a door and hatch were inoperable at the some time. ITS 3.6.14 includes a note-which permits separate Condition entry for each personnel access door or equipment hatch and would allow both a door and hatch to be inoperable for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . This change is less restrictive than the CTS requirements, however, the change is acceptable since the time to restore the door or hatch to operable status is of sufficiently short duration and is equivalent to the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months completion time which would be provided under CTS 3.0.3.

U This change is consistent with NUREG-1431.

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v McGuire Units 1 and 2 Pahe1.-134a Supplement 63l

Na Signific=t H:z:rds C:nsid:ratien Section 3.6 - Containment Systems O

O LESS RESTRICTIVE CHANGE L.38

.The McGuire Nuclear Station is converting to the Improved Tbchnical Specifications (ITS) as outlined in NUREG-1431, " Standard Technical Specifications, Westinghouse Plants." The proposed change involves making the current Technical Specifications (CTS) less restrictive.

Below is the description of this less restrictive change and the No Significant Hazards Consideration for converston to NUREG-1431.

. CTS 3.6.5.5 octions requires on inoperable personnel access door or equipment hatch be restored operable or closed in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . CTS 3.0.3 would be applicable if more than one hatch or both a door and hatch were inoperable at the same time. ITS 3.6.14 includes a note which permits separate Condition entry for each personnel access door or equipment hatch and would allow both a door and hatch to be inoperable for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . This change is less restrictive than the CTS requirements, however, the change is acceptable since the time to restore the door or hatch to operable status is of sufficiently short duration and is equivalent to the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months completion time which would be provided under CTS 3.0.3.

This change is consistent with NUREG-1431.

O. In accordance with the criteria set forth in 10 CFR 50.92, the McGuire Nuclear Station has evaluated this proposed Technical Specifications change and determined if does not represent a significant hazards consideration. The following is provided in support of this conclusion.

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

The proposed changes allow an equipment hatch or personnel access door between upper and lower containment to be inoperable at the same time. The access door and equipement hatch are not an initiator of any analyzed event, therefore, the proposed change does not offect the probability of any analyzed accident. The LCO continues to require the hatch and access door to be operable and specifies a short completion time. Therefore, the consequences of analyzed events are not changed.

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V McGuire Units 1 and 2 Page 76M of 81M Supplement 6Bl

N3 Significant Hazards C:nsid:ratien Section 3.6 - Containment Systems L 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated 7 g The change will not physically alter the plant (no new or different type of equipment will be installed). The changes in methods governing normal plant operation are consistent with current safety analysis assumptions. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident prev!?usly evoluoted.

J

3. 'Does this change involve a significant reduction in a margin of safety?

The proposed changes allow the equipment hatch and personnel access door to be inoperable at the some time. The changes do not reduce the margin of safety since the CTS 3.0.3 requirements essentially provide the same short completton time for this condition. The TS continue to require the door and equipment hatch to be operable, therefore the proposed change does not significant!y reduce any margin of safety.

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lMcGuireUnits1and2 P ge 77M of 81M Supplement 62

i McGuire & Catawba Improved TS Review Comments ITS Section 3.6, Containment Systems 3.6.14-3 JFD Bases 4 '

- STS B3.6.17 Bases - APPLICABLE SAFETY ANALYSES ITS B3.6.14 Bases - APPLICABLE SAFETY ANALYSES -

ITS B3.6.14 Bases - APPLICABLE SAFETY ANALYSES makes a number of changes to the

. second paragraph of STS B3.6.17 Bases - APPLICABLE SAFETY ANALYSES be. sed on the t - plant specific design. The changes made to Catawba ITS Bases are identical to the McGuire

. ;ITS Bases changes except in one spot. In the serond sentence of the second paragraph, the STS states the following: "...inoperability of one train in both Containment Spray System..."

in McGuire the word "both" is deleted while in Catawba it is retained. . Comment: Correct this discrepancy between the plants or provide additional discussion to justify this difference.

DEC Response:

The Catawba Bases has been corrected. The sentence is referring to a total of three systems, therefore, the word "both" is incorrect.

REVISED RESPONSE:

No changes are required for McGuire.

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mc4_cr_3.6- 3.6-58 July 2,1990 1

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h McGuire & Catawba Improved TS Review Comments

! - ITS Section 3.6, Containment Systems

- Q) l 3.6.15 Containment Recirculation Drains 3.6.15-1 DOC M.6 CTS 4.6.5.8 (McGuire Unit 2)

The CTS markup of CTS 4.5.6.8 in McGuire Unit 2 shows a change of "once per 92 days" I being added. The change is designated as DOC M.6. This change does not seem to be associated with any ITS 3.6.15 item nor is it shown in the CTS markups of McGuire Unit 1 or Caauba Units 1 and 2 which have the same specification. The justification associated with DOC M.6 is "Not used." Comment: Correct this discrepancy and provide the appropriate discussion and justification for this change.

DEC Response:

The McGuire unit 2 CTS markup has been revised to remove this change which is not used.

REVISED RESPONSE:

No changes are required for Catawba.

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( ). mc4_cr_3.6 3.6-59 July 2, 1998

4 t

McGuire & Catawba Improved TS Review Comments ITS Section 3.6, Containment Systems O

3.6.15-2 JFD Bases 5 (McGuire)

ITS SR 3.6.15.1 STS B3.6.18 Bases LCO ITS B3.6.15 Bases LCO STS B3.6.18 Bases '- LCO states the following: "The~ refueling canal drains mush have their plugs removed and remain clear to ensure...". ITS B3.6.15 Bases LCO for McGuire changes this statement to conform to the plant specific changes made in ITS SR 3.6.15.1 to read as l

follows: "The refueling canal drain valves must be locked open and remain clear to ensure..."

This change is acceptable. Howwer, this Bases change has not been incorporated into ITS B3.6.15 Bases - LCO for Catawba. Since the same plant specific changes that were made to ITS SR 3.6.15.1 for McGuire are made to ITS SR 3.6.15.1 for Catawba, the changes made to ITS B3.6.15 Bases - LCO for McGuire should be made to ITS B3.6.15 Bases - LCO for Catawba. Comment: Revise the ITS B3.6.15 Bases - LCO for Catawba to conform to ITS B3.6.15 Bases - LCO for McGuire or provide appropriate discussion and justification to show why they should not be the same. See Comment Number 3.6.15-3.

DEC Response:

- The Bases for Catawba have been revised consistent with the CTS and the McGuire Bases.

REVISED RESPONSE:

No changes are required for McGuire.

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mc4_cr,_3.6 3.6-60 July 2, 1998

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McGuire & Catawba Improved TS Review Comments-7 ,

-ITS Section 3.6, Containment Systems 3.6.15-3 JFD Bases 5' ITS SR 3.6.15.1 STS B3.6.18 Bases - SR 3.6.18.1 ITS B3.6.15 Bases - SR 3.6.15.1 and SR 3.0.15.2

- STS B3.6.18 Bases - SR 3.6.18.1 states the following in the fourth sentence: "SR 3.6.18.1 must be performed...from MODE 5 after every filling of the canal to ensure that the plugs have.

been removed and that..." ITS B3.6.15 Bases -~ SR 3.6.15.1 and SR 3.6.15.2 for McGuire modifies this statement to conform to plant specific changes made to ITS SR 3.6.15.1 to the

. following: " ...

from MODE 5 after every filling of the canal to ensure that the valves have been locked open and that..." . The same sentence in ITS B3.6.15 Bases SR 3.6.15.1 and SR 3.6.15.2 for Catawba has not been modified; yet the same plant specific changes made to ITS SR 3.6.15.1 in McGuire have been made in ITS SR 3.6.15.1 in Catawba as well as other plant -

specific changes made to the Bases for both Catawba and McGuire that change " plugs" to

" valves." Comment: Revise the Catawba ITS B3.0.15 Bases - SR 3.6.15.1 and SR 3.6.15.2

, to conform to the McGuire Bases. See Comment Number 3.6.15-2.

DEC Response:

The Bases for Catawba' have been revised consistent with the CTS and the McGuire Bases.

.,s.

k REVISED RESPONSE:

No changes are required for McGuire.

i' a

ih

?

L L

I f:

r

'O c4.<< >.e s.e-e> e '> 2. 1998

l McGuire & Catawba Improved TS Review Comments ITS Section 3.6 Containment Systems

'd 3.6 Additionalitems 3.6-01 STS 3.6.3 Bases LCO, page B 3.6-32 ITS 3.6.3 Bases LCO, page B 3.6-19 (Catawba)

ITS 3.6.3 Bases LCO, page B 3.6-18 (McGuire)

~

The Bases LCO discussion references the UFSAR location for containment isolation valves as

~ Chapter 15 or Reference 1. The correct location is Chapter 6 or Reference 3. The Bases

' have been corrected consistent with the UFSAR.

I q

L) 1 I

.O

,Q mc4_cr_3.6 3.6-71 July 2, 1998 I-i -

t Containment.Isolaticn Valves

. B 3.6.3 4

i n BASES

'/

, J)

APPLICABLE compromising the containment boundary as long as the system SAFETY ANALYSES is operated in accordance with the subject LCO.

(continued)

The containment isolation valves satisfy Criterion 3 of )

j 10 CFR 50.36 (Ref. 2). l LC0 Containment isolation valves form a part of the containment boundary. 'The containment isolation valves' safety function is related to minimizing the loss of reactor coolant inventory and establishing the containment boundary during a DBA.

The automatic power operated isolation valves are required to have isolation times within limits and to actuate on an automatic isolation signal. The lower compartment and instrument room purge valves must be maintained sealed closed. The valves covered by this LC0 are listed along with their associated stroke times in the UFSAR (Ref. 3).

The normally closed isolation valves are considered OPERABLE when manual valves are closed, automatic valves are O de-activated and secured in their closed position, blind V~ flanges are in place, and closed systems are intact. These passive isolation valves / devices are those listed in l Reference 3.

Purge valves with resilient seals and reactor building bypass valves must meet additional leakage rate requirements. The other containment isolation valve leakage rates are addressed by LC0 3.6.1, " Containment," as Type C testing.

This LCO provides assurance that the containment isolation valves and purge valves will perform their designed safety functions to minimize the loss of reactor coolant inventory and establish the containment boundary during accidents.

APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of l radioactive material to containment. In MODES 5 and 6, the I probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES.

Therefore, the containment isolation valves are not required (continued) l McGuire Unit 1 B 3.6-18 Supplement 6 L_-___--_____

Containment Isolation Valves

, B 3.6.3 ;

l BASES O -

APh.. CABLE compromising the containment boundary as long as the system SAFETY ANALYSES is. operated in accordance with the subject LCO.

'(continued)

The containment isolation valves satisfy Criterion 3 of 10 CFR 50.36 (Ref. 2).

LC0 Containment isolation valves form a part of the containment boundary. The containment isolation valves' safety function is related to minimizing the loss of reactor coolant inventory and establishing the containment boundary during a J DBA.

The automatic power operated isolation valves are required to have isolation times within limits and to actuate on an automatic isolation signal. The lower compartment and instrument room purge valves must be maintained sealed '

closed. The valves covered by this LC0 are listed alon with their associated stroke times in the UFSAR (Ref. 3 .

The normally closed isolation valves are considered OPERABLE when manual valves are closed, automatic valves are

/* de-activated and secured in their closed position, blind A . flanges are in place, and closed systems are intact. These passive isolation valves / devices are those listed in

] Reference 3.

Purge valves with resilient seals and reactor building bypass valves must meet additional leakage rate requirements. .The other containment isolation valve leakage rates are addressed by LC0 3.6.1, " Containment," as Type C testing.

This LC0 provides assurance that the containment isolation valves and purge valves will perform their designed safety functions to minimize the loss of reactor coolant inventory and establish the containment boundary during accidents.

APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment. In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES.

Therefore, the containment isolation valves are not required

> [~T l

w/ (continued) l McGuire Unit 2 B 3.6-18 Supplement 6

l l

F .

e \

('" /

i Containment Isolation ValvesAAtmosp i g atenennaric- ira tondenser. nn a 8 3.6.3 b

BASES (continued)

LC0 Containment isolation valves fom a part of the containment boundary. The containment isolation valves' safety function is related to minimizing the loss of reactor coolant inventory and establishing the containment boundary during a h

DBA-O m c> & < d g The automatic power operated isolation valves are required *kwd to have isolation times within limits and to actuate on an rg automatic isolation signal. The (47TTMmgurge valves must bemaintenedsealedclosedf6rnave r s.s instaiseo m revent/run openingj. Eunct.A ~[bi u 1u.< alen act/ath bn an4utomatic siona1N The valves covered by this LCO are listed long with their associated stroke times in t FSAR -

(Ref. .

The normally close.d isolation valves are considered OPERABLE when manual valves are closed, automatic valves are de-activated and secured in their closed position, blind flanges are in place, and closed systems are intact. These gssiveisolationvalves/devicesarethoselistedin [

f FT~\ Purge valves with resilient seals (and(sarnmary canm@-o re.ctc.fae pdId LU bypassvalves(mustmeetadditionalleakagerate requirements. The other containment isolation valve leakage rates are addressed by LCO 3.6.1, 'Cnntairenent,' as Type C testing.

. This LCO provides assurance that the containment isolation valves and purge valves will perfom their designed safety functions to minimize the loss of reactor coolant inventory and establish the containment boundary during accidents.

APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment. In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES.

Therefore, the containment isolation valves are not required to be OPERABLE in MODE 5. The requirements for containment isolation valves during MODE 6 are addressed in LCO 3.9.4,

' Containment Penetrations.*

(continued) 200hs- B 3.6-32 Rev 1, 04/07/95 Mel dd t l

\

_ - - - - - - - - - - - _ - - - - _ _ - - - . -- - _l

I L

I McGuire & Catawba Improved TS Review Comments ITS Section 3.6, Containment Systems '

-(g 3.6 Additionalitems 3.6 CTS 4.6.3.1 DOC LAS in review comments for several sections, the NRC has identified that the deletion of requirements which are redundant to regulation or items which are identified as relocated to procedures should be classified as less restrictive changes rather than removal of detail changes. Therefore, DOCS LAS, LA11, and LA17 are deleted and replaced by DOCS L35, L36, and L37 to be consistent with other changes made to other sections and to eliminate the relocation of a requirement to a procedure.

O v

4 mc4_cr_3.6 3.6-72 July 2, 1998

$06&8th/A.bOV1 4 56*b '

CONTAINMENT SYSTBis .

. M5 KTEE2b CONTAlleeff ISOULTION VALVES OMITammm=.-IflallFtM/tlpFRATIfb 4CO 3.6.3 , M-- ta amant isolation valves shall be OPERAg8L an- miation teg _

s r= . 1m t.-

- -- = sum ----.. - . m _

b APPLICABILITY: MODE 5 1. 2. 3, and 4. *N/D 4,8 'I

<= Q % _

+ u-, m> nc@ q@'%,,

hithoneormore 16 .

., . _s. ... 4 _ inment . . ._ isolation

_ ;;_ ._.a valve . . a . .@ . . inoperablehntain

_ ,,,w atXa re the inoperable)(alve(s) to OPERABLE statA witate 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months sg

& Isolate ected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one deactivated automatic valve secured(cdeue: in the isolation posih,res.74>M M d Q

Acriset 4.1 or /,c.uAus.

Isolate O affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least t.p< . fee.see 1 1

.g A '" # N d'2. one closed manual valve or blind flange, or,c ,c l- hNT # @ Be in at least NOT STAH08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDonal within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

gg,g g :

e. The prov fens of Specification 3.0.4 a not applicable provided '

sk7tcN S that affected penetration is (sel in accoNance with ACT

b. or ..above. and provided that th associated system, if app - h for to cab , is declared inoperable and fe that svstas are performed. f appropriate ACTICII stat ts; SURVEILLANT Z i~ eis O (4.6.3.1 -.ainment isolation valve u se sensestrates artunsi.r. or' <

. to returni valve to service after atenance, repair er replacemen wort is reed en the valve or its as lated actuator, control er r-kircuit at:rformance of a cycline t and verification of isolation me.

.# gj yed,, >,4 u # A ..o

.GC_._3 tr. 5. ...

E3th11selatten valveash411 Wstrated OPDIABLE LTWMaai bss M g ggyv== mr ===> nm ===>at Teast page per 18 mosths bq s

" j t en a Phase A Centat -_ . asosauen test sayisi de**f d *

{ g N;f, g,g. vertryt Phase A lation valve actuates its isolation position,

=f 8'5 d'*"

l

b. Verify Pha that en a Phase B Ces imment isolation test si 1 ead 8 selatten valve actua s te its isolatten positi , and '

T ,g b ,,, u l

c. Ve fylag that en aiantaEa-at madinardvie rmigW signef, adu4 /er-4,, g ( and exhaust valve a;*=tes to its isolation itian- 5,.,,14/,./,

St 6,g.~ The isolaties tim,e tobe ,ithin e ii.iof eachC-r m_._ a=r=+ama autenatilvalve

_.- -- , _ - . . . . - . - mi )

n. @ _wa a a.e.

ue na rn a.,

i.1 -

ficGUIRE - tulli 1 3/4 6 16 Amendment No. 166 l

~

O 1

Pge /#f tz.

a Spa Sea.6mn 5.67 0 J, (, CONTAllMENT SYSTDt5

'e 3,4,7 6NYDROCEN RfC(MBINER$

dINITINFsXUICITION F00 OPERAUGG 4CO J/,.7 (Z" IPA Two smernennet ranrainmen Hydrogen Recombineg@shall be OPERASLE. I APPLICAfnLITY: MODES I and 2.

ACTION: INN I1 ANN 2 -

[sfRD (ne Hydrogen Recombiner EiiDen inoperable, restore the (manerable system to OPERABLE status within 30 dayslor be in at least HOT STANDSY within the Ar,q meat t nowrs.

Supvf!LLAlICE RE0pIRO4ENTS

@ Each Hydrogen Rec shall he '8-<trated OPERASLE:

qa w .,) ..

54 5,6.7./ Q At least once per mont s by wiyog euring/ combiner System functional test i ii wisimum heater snesta t=in- .m. -

reater snan equal to 700*F within 90 minutes. Upon reachingurrl .

(togF, 700 increase he power settlog to maximum power or 2 minutes -

verify that namer meter made mesatse than er analto60kW]and W

@ At least once per 18 months by:

m ~ Performing a CHANNEL CALIBRATI0fl of all recombiner Tinstnamentath and raaten1 circuits._r--

4 i ,j r g 3,4* 7, g @ Verifying through a visual examination that there is no evidence of abnormal conditions within the recombiners enclosureft.e.,a nn=3 or strucuras_sennecuoasg- o eemosus of faceta=

art =1= *te.l . andr GCU -

43.4.7.3 8 **'r' 'a *** iat=='ft' '8" *****r *ctr'c'

- c'"t5 60 performing a resistance,to y testye m; _ ..

t+, g QIr17MVareater +1er ^1 * *a 1

I 1

McGUIRE . UNIT 1 3/4 6 18 Amendment No. 166 !

i O ar <s 2 t

Ger:bczkrs.3.foJf,e i

4 CONTAINMENT SYSTEMS

@ REACTOR BUILDING FIT NT d_i d co d ON FOR IRATIOD -

g,f 4 f,0 3.6 kkNBLE)

, The itsvefurakflite_ ark 6 efAhlreactor' building shall bg,Fajittai cut yieveJA:enstsJent unpe es.gytancej,riteriap spe53.ncattg4.6 ..

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

Ell 9!!: ui o d '.

\

A @ Mre,'7 M"'UN*' M Mii M *itE* M t*"*)

g, , G3sPttr i p ung sn3 meettor Copystesy . ... sun: as py

/W W MSO $Wn aY Apak SURVEILLANCE REOUTREMENTS fiftstrCIS SA A L /4'l .

se S.3,y, 3 du The stavetural integrity of the reactor building shall be deterstned !

ng the shutdown for each Type A containment leakage rate test specification.KEP?M by a visual inspection ofible the exposed /om ac

@45 4.1./h interior M=ie-g, i ande exterior - ; in -warance surfacesofofuthe reactor butiding/anapertfying dAfD ddradattad.lf the T -n.wse surfaces o( othef abn 1 i g fg A t is p ormed Kt 10-ysar laMrvay, two addit 4onallpspection hall e perf at/pproxisfately/equavinte is duri/H shu+ h_= b+ T A tes ctd buildt detected during the y w Io rasa on o sne t A reporfedto ve vtg red i t ss 11 bet ,

Commis9 en puryGant 4 10 CFR tio 50. .a 50.73.) fJS '

v .

McGUIRE - UNIT 1 3/4 6-10 Amendment No.173 I

p e /p;f .5 V

I l

1

4

. sw .<. a . ,, e.a.,

ConrAimeur sysTes 5.45 O (GCUD carrAlmDrf IsosaT10N VALVES (i_ininum ~ !Tiedi PDR ci~=UdiD b zco

.3.6.3 _ . @iaseet

_ .isolatten

_ _ _ . valves

. _ . shall

. _ . ,be _ .OPetA8tifwith

, 1selatiesvtimes7.}

canrs m- -ru m - - i...s. d..o s a , n,ae mg w& A w-y Me '~+

p.m ene e ,s. au ..c e.retc.awas.st

. _ a = ... _is.1stien vaive@ sa.,erahie gassa es mast 1. - .

atje (n. wre the inoperable valve (s) to uruuuu szasus within s 6' Isolate effectedpenetrathwithin4hoursgathposittan: use of at least A alag A.i one deacti automatic valve secured (a the iso or c @ ,4 , a ,,A (4

}

h Isolate ected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least]h one closed mensal valve er blind flange, irr ,

L A"'## d. Se la at least NOT STANDSY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD Acts,# 4 2. F y SHirTDolet withis the folleming 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d## I/ The p isions of Specift sa 3.0.4 are not app cable proviees g,,,y p that affected is isolated in acc with ACTI

b. c. above p that.the assecta system. if app -

umr f cab . is decla. red i le and the assrom te ACTien sta tsj g,a g I that system are reed. f 68f to navrve e aser asastseens 04.3.1- contatament isolation vaM 11 he demonstrated OPGtASLE o to aturni the valve to service after as . repair or replac 5 wert is en the valve or its as isted actuator, control er r if (circuit performance af a eveller test verifirmeine of isolatt tee ,

) $Jt JG ~ _ %g.30 Un sha a trated OPDtAtt.E N /'***

,' }

[h*

~

}

at,Ieastonceperlaasethsb( ,

. v.+. . a . -, ov e . - ,, -

== ties ,ai . .ct t. to is.iati . ,. sis w,s. . -

b. Verifyin en a Phase 5 Cental Isolaties test signa each ed I

^

g,g Phase O ' lation valve actuates its isolation posittoa. g I. Verl ,that emba='*1====t "% act' vinv 4ttah test si 1. eac 74st:4n eri t

p. " t valve acteates'to ' ts se' ation sositi 4

en ac6usi

43.4.5 -

4232D Thetoiso be 14 tion t.ietu.i withia of.ea.ch eewer

_ _ _ emeratage_.estematt&

.7......- valve % (

, peroaaAnn e. A l'Y que nmmeAh'9 f&

Mc4NIRE - tAIIT 2 3/4 6 16 . Amendment No. las O

2p. Ishz.

I pedadeA S.G. ~/

i g

u ,mm

3. t. 7 tsisirin-- -- -

inna - mm a eraria uo s.< 7 ccov M % id e

- n AprLic4srt m Imots 1 d 2.

gag: //~Mer rta '

,,,,y t o g ..... , =.=.,i i..

.......f_...

i i. .,.

Lf 0 TlW b -

,,"""'If f marr m.s.gMT. ,

a s.4 7. / -

< m 5.a ~ - i

. .o t ,

.,, yng =====

-.,t

"_"_'r;,;;r sw, a.si. --

q..

j L""."l.c.e th.

fy th.t he --

p.m.rJftti.g

^

t ste p#r for te.d. u.t.c th .r t.al t.d 80 le gf

% At 1 t p.r la eenths by:

O+2% = = ,,,rJ " - - > .

,3 x3.c.st an ,i. g,i . i.

u t - ..

~'~~ sic:a- ze.c,. z; = 8 ru.t..u 21 y,gg;ggggT:!;;, l 6'J".*..t7.i::'.',"l%* .

g y[U".#tdarGe@

~

. steestM 6 W IT t 3/4 6-14 A dm t 5.". 148 O -

dp / of .2.

r j

spa;tialaa r.us

. h CONTAINMENT SYSTEMS h AEACTOR BUILDING STRUCTURAL INTEGRIT (LIERfM COMPfff0N MEM

~

Q

t,co COAw6/r3 3.6&D The disuttuckTatMt CatyteveLesisteps unap -4 reactor building shall bef ta,1 l -p=Sce,Antepn 5y ynica 4,6; .

APPLICABftffY_: leDES 1, 2, 3, and 4.

ACIEE:

jgg ec Ar new I sra n[s. '

W'"5 eJ- - -

rMr$7edNr" U

. p to sp...n asp =MstM [nts t$ M

.u vr m santj ystem t =j-rature abye .vv c. f 4MgW N [ &

IC Q%SS&W SURVEILLANCE REQUIREMENTS 9 {iMMk&5R v n ./4.

54./6a/

~

the shutdown for each Type A containment leakage )

0*/' >-taterior Speci and(cationce.zwF.D by a buildin exter or surfaces of the reactor visual inspection ible of* the * *d hse 5

6, - rdati. .

m ==c==Wom pfifyi

- - ersaces oyother _ _ -rual $0 the Type A test is performed at 10-year intervals, twar l'aostttonal inspections shall be performed at approximately equal35intervals durine shutdowns between Type A testss any1.. .

Yreacteristising eetecten ounng the above requ re i .. r.o.uen e1 6a t rted to tiie Consission pursuant to 10 CFR d inspections shall bel o .,

ions 50.72,and50.73g McGUIRE - UNIT 2 Amendaent No. 155 O

ge M J

, Discussitn cf Changes

. S:ction 3.6 - C'intainment Systems (n) TECHNICAL CHANGES - LESS RESTRICTIVE only eliminates a duplication of requirements and the CFR provides sufficient regulatory control over this activity. This change is consistent with NUREG-1431.

L.34 CTS SR 4.6.1.1.a requires each penetration located outside containment to be verified closed by a volve, blind flange, or deactivated automatic valve once per 31 days. ITS SR 3.6.3.3 excludes penetrations that are locked, sealed or otherwise secured in position from this verification. This change is considered acceptable since any penetration that is secured in the correct position is meeting the safety function. These penetrations would not be inadvertently mispositioned since they are secured in the closed position. This change is consistent with CTS exceptions provided for other volve alignments, e.g.

ECCS, which are locked, sealed, or otherwise secured in position. This change is also consistent with NUREG-1431.

L.35 CTS 4.6.3.1 Surveillance Requirement requires on operability verification for each containment isolation valve prior to F' returning the valve to service after maintenance, repair or G) replacement work on the volve or its associated actuator, or control or power circuit. These details describing when post maintenance testing is required to be performed is redundant to controls in the QA Topical Report and is not necessary for inclusion within the TS. Changes to the QA Topical Report are evaluated under the requirements of 10 CFR 50.54. Any change, using this criteria, will ensure proper review and conformance to the QA Program requirements. This change is consistent with NUREG-1431.

L.36 The CTS 4.6.4.2.b.1 requirement for a channel calibration of the hydrogen recombiner is deleted. The hydrogen recombiner instrumentation does not relate directly to the system OPERABILITY and is not necessary for inclusion within the TS. Control of the ovatIability of, and necessary compensatory activities if not available, for indication instruments, monitoring instruments, and alarms are addressed by plant operational procedures and policies which are controlled by the plant procedure control program. In addition, the system functional test required by ITS SR 3.6.7.1 will ensure that necessary controls will functton properly. This p change is consistent with NUREG-1431.

V McGuire Units 1 and 2 Page L - 1242 Supplement 63 l

Discussien cf Changes

.' S;cti:n 3.6 - C*ntainment Systems TECHNICAL CHANGES - LESS RESTRICTIVE L.37 The CTS 4.6.1.7 requirement to report abnormal degradation discovered during reactor building surveillance is deleted. The change is acceptable since inspections and reporting of degradation are required by 10 CFR 50.550 and are not necessary for inclusion within the TS. Requirements associated with regulations are not necessary to be duplicated in the TS and are implemented directly. . These changes are consistent with NUREG-1431.

L.38 CTS 3.6.5.5 actions requires an inoperable per:wnnel access door or equipment hatch be restored operable or closea' in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . CTS 3.0.3 would be applicable if more than one hatch or both a door and hatch were inoperable at the same time. ITS 3.6.14 includes a note which permits separate Condition entry for each personnel access door or equipment hatch and would allow both a door and hatch to be inoperable for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . This change is less restrictive than the CTS requirements, however, the change is acceptable since the time to restore the door or hatch to operable status is of sufficiently short duration and is equivalent to the

) 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months completion time which would be provided under CTS 3.0.3.

This change is consistent with NUREG-1431.

i l

/3 ,

U l l

McGuire Units 1 and 2 Page L - 1313 Supplement 63l l L_______ __

Discussion of Ching2s

. 'S cticn 3.6 - Containment Systems

-(

( TECHNICAL CHANGES - REMOVAL OF DETAILS program required by ITS 5.0, Administrative Controls. Any change, using this criteria, will ensure proper review and conformance to the_ requirements of 10 CFR 50, Appendix J. This change is consistent with NUREG-1431.

LA.4 CTS Surveillance 4.6.1.3.d contains detail information concerning the pneumatic seal system for the personnel air lock. doors.

' Testing of airlock door seals is required by 10 CFR 50, Appendix J, Option A. The detail information contained in the CTS Surveillance, including any exemptions to 110 CFR 50, Appendix J, is relocated to the Bases for ITS 3.6.2. The movement of this information'is appropriate because it involves details that are not necessary for inclusion in the LCOs and are more appropriate for the Bases. Any changes to the Bases are evaluated under the Bases' cont'rol program required by ITS 5.0, Administrative Controls. ITS SR 3.6'.2.1 requires that airlock testing be conducted in accordance with 10 CFR 50, Appendix J, Option A and provides the acceptance criteria. 'This change is consistent with NUREG-1431.-

. LA.5 -Not used. CTS t.5.3.1 Surv ill:::: R:;uir; :nt r ;uire: :n

per:bilit; ::rific:ti:n f:r :::h :::t:in:::t i::1:ti:n ::lve prier to returning the valve te ervice after . inten:nce, repair

..r. _____._____i......._.__o......t.

.. m. i. m. . __

. .. 4... ..__,..

ntr:1 er p _ r circuit. Th::: det:il: d:: ribin; when :n SS i:

7:q; ired t: he performed i: ::r: :ppr:prist f r pl:nt prc :dare:.

":Ving thi; i-f:rt:ti:n te pr ::d r : i: n:t On imp :t On th:

. . . t s. ,. ._

r.. t.._ ,... t. .

. .-.2

. , . . , . . .t._._..__.. 4......i.,m __i.

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

- dir::ti:n th:t i; implicit in the an Pr:gre: centrel: :::::i ted with th: :: int:::r.:: pr:::::. Any ch:nge; t; the prc::d rt: Orc

" 1;;t-d under thc.pr^::d r :: tr:1 pr^;r :. Any th:n; , ::in;

.t..e.,..__.,..__,.......

,s.i.

. . . . . . . ________........;___,___.___..un,,.,

r..r., ...... .... ............... .. .-.

Pr:gre: 7:q;irc :nt . Thi; th:ng i: : n:i: tent with TJREC lt31.

t LA.6 The descriptive . material contained in the CTS 4.6.3.2 regarding the details of which test signal'(Phase A, Phase E, or Safety L Injection) should be applied to isolation valve testing is moved

~

ll to the Bases for ITS 3.6.3. The Containment Radioactivity-High

-\

L McGuire Units 1 and 2 Page LA - 28 Supplement 63 l L_-_-_-____-_--_-__-____-_____-__-__________-___________________________________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .

Discussitn af Chrnges

.' SIcticn 3.6 - C:ntainment Systems C

\v\/ TECHNICAL CHANGES'- REMOVAL OF DETAILS for the Bases. The ITS Bases are subject to the controls described in ITS Chapter 5 " Administrative Controls." Changes to the Bases are evaluated under the 10 CFR 50.59 criteria. Any change, using this criteria, will ensure proper review. This change is consistent with NUREG-1431.

LA.11 Not used.The CTS 4.5.1.2.b.1 requirement f;r ; channel calibration Of the hydr: gen rec; biner i; mcVed tc plant procedurc . The hydr: gen rec 0=biner instrumentation dec: not necc :arily relate directly to the y te: OPERABILITY. Control of the avail:bility ef, and nece;;;ry c0= pen::tery activitic; if not available, fer indication in:trument , Onitoring in:truments, and clarm; cre ddrc :cd by plant Operational procedure; and policic; which cre c0ntrolled by the pl:nt precedure centrol program. In addition, the y;tc= functional tc:t required by ITS SR 3.5.7.1 cill en;ure that necc : ry centrol: will function properly.

LA.12 CTS 4.6.5.6.1.a and d identify the test signal for starting the air return and hydrogen skimmer fans. This type of descriptive

' / information is not normally included in the specification but is

\~ better presented in the Bases. Therefore, this detail is moved to the ITS 3.6.8 and 3.6.11 Bases. The Bases are subject to the controls described in ITS Chapter 5 " Administrative Controls."

Changes to the Bases are evaluated under the 10 CFR 50.59 criteria. Any change, using this criteria, will ensure proper review. This change is consistent with NUREG-1431.

LA.13 Descriptive information regarding the Annulus Ventilation System {

surveillance in CTS 4.6.1.8 and Air Return System in CTS !

4.6.5.6.1 is moved to the ITS 3.6.10 and 3.6.11 Bases, respectively. The movement of this information is appropriate because it involves details that are not necessary for inclusion in the LCOs and are more appropriate for the Bases. The Bases are subject to the controls described in ITS Chapter 5 " Administrative Controls. " Changes to the Bases are evaluated under the 10 CFR 50.59 criteria. Any change, using this criteria, will ensure

proper review. These changes are consistent with NUREG-1431.

i i

l v

I McGuire Units 1 and 2 Page LA - 48 Supplement 63 l ;

4 i

Discussion of Changes

. Secticn 3.6 - C:ntninment Systems

/m h TECHNICAL CHANGES - REMOVAL OF DETAILS LA.14 Descriptive infonnation regarding the Ice Bed LC0 and surveillance in CTS 3.6.5.1 and 4.6.5.1 is moved to the ITS 3.6.12 Bases. The movement of this information is appropriate because it involves aetails that are not necessary for inclusion in the LCOs and are more appropriate for the Bases. The Bases are subject to the controls described in ITS Chapter 5 " Administrative Controls. " Changes to the Bases are evaiuated under the 10 CFR 50.59 criteria. Any change, using this criteria, will ensure proper review. These changes are consistent with NUREG-1431.

LA.15 Descriptive information regarding the Ice Condenser Doors surveillance in CTS 4.6.5.3.1 and 4.6.5.3.2 is moved to the ITS 3.6.13 Bases. The movement of this information is appropriate because it involves details that are not necessary for inclusion in the LCOs and are more appropriate for the Bases. The Bases are subject to the controls described in ITS Chapter 5 " Administrative Controls." Changes to the Bases are evaluated under the 10 CFR 50.59 criteria. Any change, using this criteria, will . ensure proper review. These changes are consistent with NUREG-1431.

I \

V LA.16 Descriptive information regarding the reactor building surveillance in CTS 4.6.1.7 is moved to the ITS 3.6.16 Bases.

The movement of this information is appropriate because it involves details that are not necessary for inclusion in the LCOs and are more appropriate for the Bases. 1he Bases are subject to the controls described in ITS Chapter 5 " Administrative Controls."

Changes to the Bases are evaluated under the 10 CFR 50.59 criteria. Any change, using this criteria, will ensure proper review. These changes are consistent with NUREG-1431.

LA.17 Not used. Descriptive infc = tion regarding the repcrting of abac= 1 degrad: tion discovered during reacter bui44hg surveillance; in CTS .5.;.7 i: =cved 10 the plant procedures.

The movement of tM: infc = tion i appropriate because it i r,vcive repcrt; required by 10 CFR 50.72 :nd 50.73. Requiremer,t l awee4:ted with regulation are not necc :ary tc be duplicated in l et TS :nd arc implemer,ted directly inte precedure:. Change; to I plant precedurc; are ubject to the site prccedure contr:1 program

, which en;ure; prcper review and confc= nce with regulatcry

' requi rement;. The:c-change are cen;ister,t with N'JREC 1431.

r3 V t

V r

lMcGuireUnits1and2 Page LA - 58 Supplement 63

o N3 Significant Hazards C:nsideratien

, .' S2cticn 3.6 - C:ntsinnent Systems

- LESS RESTRICTIVE CNANGE L.35 The McGuire Nuclear Station is converting to the Improved Technical Specifications (ITS) os outlined in NUREG-1431, " Standard Technical

. Specifications, Westinghouse Plants." The proposed change involves making the current Technical Specifications (CTS) less restrictive.

Below is the description of this less restrictive change and the No Significant Hazards Consideration for ~ conversion to NUREG-1431.

CTS 4.6.3.1 SurveilIonce Requirement requires an operability verification for each containment isolation valve prior to

. returning the volve to service after maintenance, repair or replacement work on the volve or its associated actuator, or control or power circuit. .These details describing when post maintenance. testing is required to be performed is redundant to controls in the QA Topical Report and is not necessary for

' inclusion within the TS. Changes to the QA Topical Report are evaluated under the requirements of 10 CFR 50.54. Any change, using this criteria, will ensure proper review and conformance to.

the QA Program requirements. This change is consistent with NUREG-1431.

+

In accordance with the briteria set forth in 10 CFR 50.92;'the McGuire Nuclear Station has evaluated this proposed Technical Specifications change and determined it does not represent a significant. hazards .

consideration. The following is provided in support of this conclusion.

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

The proposed change removes the surveillance requirements for

^

. post-modificotton and post maintenance testing of containment isolation valves.- The requirement to perform appropriate post-modification or maintenance testing of systems continues to be applicable to any modification or maintenance activity in the facility as required by the QA Topical Report. Since no substantive change in the plant controls applicable to post-modification or maintenance testing is proposed, the change does not involve a significant -increase in'the probability or consequences of an accident previously evaluated.

I u

O m ,,e units t .n o ,.,ez ot,m S o ement ,o 1

I N) Significant X:zards Consid:raticn

.' S:ctitn 3.6 - C:ntainment Systems

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

The proposed change does not involve the installation or use of any new or different equipment. Nor does it involve any new or different mode of operation of the plant. The proposed change places the administrative control for post-modification and maintenance testing with the QA Topical Report controls established for all other post-modification testing. Therefore, the change does not create the possibility of a new or different kind of accident than previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

Existing QA Topical Report controls based on the regulations and applicable standards provide adequate assurance for maintaining the appropriate margin of safety as a result of plant '

modifications and maintenance activities. A part of these controls are the application of post-modification and maintenance testing criteria to ensure the OPERABILITY of systems prior to

( their return to service. Plant systems other than containment isolation volves, with equal importance to overall plant safety are adequately tested with these existing controls. The removal of the explicit post-modification surveillance requirement and transfer of the responsibility for this testing to the existing QA Topical Report controls is consistent with the margin of safety provided for changes to other systems important to safety.

Therefore, this change does not . involve a significant reduction in a margin of safety.

?

3 (J \

l lMcGuireUnits1and2 Page 7Ha of 7973 Supplement 63

N3 Significat Nutrds Consid rction 1 S;ctien 3.6 - C:ntoineent Systems ;

.LESS RESTRICTIVE CHANGE L.36 The McGuire Nuclear Station is converting to the improved Technical Specifications (ITS) as outlined in NUREG-1431, ' Standard Technical Specifications, Westinghouse Plants." The proposed change involves making the current Technical Specifications (CTS) less restrictive.

Below.is the description of this less restrictive change and the No Significant Hazards Consideration for' conversion to NUREG-1431.

The CTS 4.6.4.2.b.1 requirement for a channel calibration of the hydrogen recombiner is deleted. The hydrogen recombiner instrumentation does not relate directly to the system OPERABILITY and is not necessary for inclusion within the TS. Control of the availability of, and necessary corppensatory activities if not available, for indication instruments,~ monitoring instruments, and alarms are addressed by plant operational procedures and policies which are controlled by the plant procedure control program. In addition, the system functional test required by ITS SR 3.6.7.1

'wili ensure that necessary controls wili functton properly. This change is consistent with NUREG-1431.

O In accordance with the criterta set forth in 10 2FR 50.92, the McGuire Nuclear Station has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards considerotton. The folIowing is provided in support of this conclusion.

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

The proposed change removes the requirement from the Technical Specifications ^ to perform calibrations on the hydrogen recombiner instrumentation. The instrumentation is not assumed to be the initiator of any analyzed event,,therefore, the change does not offect' the probability of an event occurring. The proposed change does not affect the functional ability of the system to perform during an analyzed event. The functional' ability of the recombiner is adequately determined by required functional testing. Therefore, the proposed change does not affect the consequences of a previously analyzed event.

L O'

-McGuire Units 1 and 2 Page 7274 of 7973 Supplement 63l

- __ ________ - __-_ _____ = ____ - _

1 N3 Significant H:z:rds C:nsid2ratirn 1 l S:cticn 3.6 - C:ntainment Systems G' 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve the installation or operation of any new or different kinds of equipment. Nor does it involve a i new or different mode of operatton. The proposed changes do not result in systems operating in a manner different from existing procedure < and practices. . Therefore, the proposed change does not I create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of

. sofety?

The proposed change removes the requirement from the Technical Specifications to perform calibrations on the hydrogen recombiner instrumentation. The system does not specifically require this instrumentation to be used to meet the required functional capability. The verification of functional capability is determined by required functional testing, which includes system controls. Therefore, the change does not involve a significant reductton in a margin of safety.

O lMcGuireUnits1and2 Page 73M of 79M Supplement 62 l

_ - - - - - )

N3 Signific nt H:zards Ccnsid;rcticn l S:ctign 3.6 - C:ntninment Systems

/]

Q LESS RESTRICTIVE CHANGE L.37 I

l The McGuire Nuclear Station is converting to the Improved Technical Specifications (ITS) as outlined in NUREG-1431, " Standard Technical Specifications, Westinghouse Plants." The proposed change involves making the current Technical Specifications (CTS) less restrictive.

Below is the description of this less restrictive change and the No Significant Hazards Consideration for conversion to NUREG-1431.

The CTS 4.6.1.7 requirement to report abnormal degradation discovered during reactor building surveillance is deleted. The change is acceptable since inspections and reporting of degradation are required by 10 CFR 50.55a and are not necessary for inclusion within the TS. Requirements associated with regulations are not necessary to be duplicated in the TS and are implemented directly. ' These changes are consistent with NUREG-1431.

In accordance with the criteria set forth in 10 CFR 50.92, the McGuire Nuclear Station has evaluated this proposed Technical Specifications change and determined if does not represent a significant hazards

0) %

consideration. The following is provided in support of this conclusion.

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

The proposed changes delete requirements from Technical Specifications which are already addressed by CFR requirements.

The location of regulatory contrpls is not on initiator of any analyzed event, therefore, the proposed change does not affect the probability of any' analyzed accident. The CFR is directly implemented and enforceable. Therefore, the requirements associated with the CFR will continue to be met. The safety analysis assumptions associated with analyzed events are not affected by the source location of regulatory requirements, therefore, the consequences of analyzed events are not changed.

1 (3

~L) I l

McGuire Units 1 and 2 Page 74 M of 79 M Supplement 63l j

N3 Signific'nt H:zards C nsid;raticn 1 S:ctien 3.6 - C:ntainment Systems L

/

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

The change wiil not physicolly alter the plant (no new or different type of equipment will be instalied). The changes in methods governing normal plant operation are consistent with current safety analysis assumptions. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The proposed changes delete requirements from Technical Specifications which are adequately addressed in the CFR. The changes do not reduce the margin of safety since the CFR is directly implemented and enforceable and the location of regulatory requirements has no impact on any safety analysis assumptions.

O b

V l McGuire Units 1 and 2 Page 75M of 79M Supplement 63

4 McGuire & Catawba Improved TS Review Comments ITS Section 3.6, Containment Systems l

V l

3.6 Additionalitems 3.6-02 CTS 4.6.5.6.2 (McGuire only) ITS SR 3.6.11.7 DOC A22 1

The CTS requires that the operation of the air return fans function within the setpoints for the containment pressure control system. The ITS provides the expicit functional requirements for

' the air return system in response to the CPCS setpoints, similar to the CTS requirements for Catawba. ITS SR 3.6.11.7 requires the air retum fan dampers to close upon receipt of a terminate signal. This signal will prevent the damper from opening, but does not cause the damper to close. The ITS SR is revised accordingly to delete any reference that the terminate signal will close the valve.

O mc4_cr_3.6 3.6-73 July 2, 1998

l ARS

," 3.6.11 SURVEILLANCE REQUIREMENTS (continued)

\

SURVEILLANCE FREQUENCY 1 SR 3.6.11.2 Verify, with the ARS fan damper closed and 92 days with the bypass dampers open, each ARS fan motor current is s 32.0 amps when the fan speed is'a 840 rpm and s 900 rpm.

SR 3.6.11.3 l' Verify, with the ARS fan not operating, 92 days each ARS motor operated damper opens automatically on an actual or simulated j actuation signal after a delay of 2 9 ,

seconds and s 11 seconds.

SR 3.6.11.4 Verify the check damper is open with the 92 days air return fan operating.

\

v SR 3.6.11.5 Verify the check damper is closed with the 92 days air return fan not operating.

SR 3.6.11.6 Verify that each ARS fan is de-energized or 18 months is prevented from starting upon receipt of a terminate signal and is al. lowed to start upon receipt of a start permissive from the Containment Pressure Control System (CPCS).

SR 3.6.11.7 Verify that ARS fan motor-operated damper 18 months is allowed to open upon receipt of a start l l permissive from the Containment Pressure Control System (CPCS) and is prevented from )

opening in the absence of a start pennissive.

l r

'O V

McGuire Unit 1 3.6-29 Supplement 6 l l

l

ARS

3.6.11 SURVEILLANCE REQUIREMENTS (ccatinued)

SURVEILLANCE FREQUENCY SR 3.6.11.2 Verify, with the ARS fan damper closed and 92 days with the bypass dampers open, each ARS fan

. motor current is s 32.0 amps when the fan i speed is a 840 rpm and s 900 rpm.

SR 3.6.11.3 Verify, with the ARS fan not operating, 92 days each ARS motor operated damper opens automatically on an actual or simulated actuation signal after a delay of a 9 seconds and s 11 seconds.

1 SR 3.6.11.4 Verify the check damper is open with the 92 days air return fan operating.

O SR 3.6.11.5 Verify the check damper is closed with the air return fan not operating.

92 days SRf 3.6 11.6 Verify that each ARS fan is de-energized or 18 months is prevented from starting upon receipt of a terminate signal and is al. lowed to start upon receipt of a start permissive from the Containment Pressure Control System (CPCS).

-SR 3.6.11.7. Verify that ARS fan motor-operated damper 18-months is allowed to open upon receipt of a start l permissive from the Containment Pressure Control System (CPCS) and is-prevented from opening in the absence of a start permissive.

I O

,McGuire Unit 2 3.6-29 Supplement 6 l

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

f INSERT SURVEILLANCE FREQUENCY SR 3.6.11.4 Verify the check damper is open with 92 days the ARS fan operating.

SR 3.6.11.5 Verify the check damper is closed with 92 days the ARS fan not operating.

SR 3.6.11.6 Verify that each ARS fan is 18 months de-enercized or is prevented from startin; upon receipt of a terminate signal and is allowed to start upon receipt of a start permissive from the Containment Pressure Control System (CPCS).

SR 3.6.11.7 Verify that ARS fan motor-operated 18 months

/' damper is allowed to open upon receipt i of a start permissive from the Containment Pressure Control System (CPCS) and is prevented from opening i in the absence of a start permissive. l Insert Page 3.6-52 McGuire

_ -- )

f

'O ENCLOSURE 4 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION SECTION 3.9 O

l' f 4

McGuire and Catawba improved TS Review Comments O

-Q ITS Section 3.9, Refueling Operations l

3.9.1-02 DOC A.1 CTS 3/4.9.1 Applicability Note

I Deletion of the footnote qualifying Mode 6 Applicability referred to as A.1. This change modifies the CTS requirement for Mode 6 applicability. This is not an Administrative change and A.1 does not address the deletion. Comment: Provide an appropriate justification and DOC.

~ DEC Response:

DEC disagrees. The footnote is simply a repeat of the definition of MODE and MODE 6 as found in ITS section 1.0 and does not represent any change in technical requirement. ITS Table 1.1-1 defines MODE 6 in part as ..."c) One or more reactor vessel head closure bolts less than fully tensioned."' ITS Section 1.0 defines MODE in part as ..."with fuel in the reactor vessel." It is unecessary and inconsistent with the writers guide to repeat the definition of a defined term. Therefore, the deletion of this redundarit information is a format only change, DOC A1.

l REVISED RESPONSE:

l

~ l DOC A11 is added to provide the above justification to the CTS markup.

l 1

i L mc3_cr_3.9 4 May 21,1998

. Specid a hen 3. 9./ -

O- *; - a REFilEI.ING OPERATION 5'

= - - -

,e w CLIstItIng cosmLTION FOR OPEndTION) i

@a 5.S$4252P GS/oma concentrate e n is=a nar== of the Iteactor Coolant J System, ant the refueltag cana17shall be maintainedynuors ano sumciesrc to)

OA.t s W... C. .G,Mk - - - ruscuve s% 4=utWufred av uwmroi ene. sowing com reaftivity conditia=( ic/

ra. Eithjfr a Q of 0.95p less, or X r os s e i ra L t R APPLICAREL TY: MODE 6', with tar reactor ve<<=i ==== closure bolts less snarp r_ui nn -- ones or wm the heat ._. 4f

.I

ACT!001: ,

A W'ith the requimments of the above specification not satisfied.-immediately sospend all operations involving CORE ALTDATIONS or positive reactivity'

. chaeges_and initiate pna 6- u..-, .u 5 ,, m wr tnan or equal to gps c . . uen m ing greater than or al to 7000 ppe boren or l equivalent until is reduced to les or equal to 0.95 or t boron concentration is tored to greater an er equal to the minimum oren '

44 1 concentration ified ia 'ha fa" ^= "ma limi*< ^^ 44 ===" i n th*l more restrict eg fg% go,.,,,g y,4 m g g g Lfowihm /m/s.f g O - " " - - " - "

.t.1.1 The restrictive of abbvetworwectivityconditi 's shall he]

termined p or to

a. og or unbolting reactor vessel head, and

. b. Withdrawal of a 1 length control rod in a s of 3 feet f its fully inser ed ition within the reactor ssel. r

@ 88 / /) gX.T2) The borou concentration er en* =aktar -i-r svstem man ene rarmaia sspzgg3 mu - _. .-- - - --.cmi --i,=.=,at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

& fxasm +sek st spa (wh #e couO b* 94.g.1.3 M-250 shall be verified closee sneer aantaistrative control at leas M

D'S M*" '"'(('#'

M#j" esce 132 erper 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months w -1026 M 131 M-140, M-176, M-468 M-808, and either M-shallor$e verified closed under administrative control at le J.9.2 , %borded esce per 12 howes when accessey to makeup to the EWST during refueline u/aler 4xrcc.. sperations.f isolsfr5 UlUr'W T*The 6 tdienever fuel 1s in zne reactoh vessel shall be maintainee the vessel head cleenin[ halts 1 << **ma fully tegioned or with? r the head removed. f p sicGUIRE - IN11T 1 3/4 9-1 Amendment No. 166 O~ .

Pne. W /

o

Sp-c,/;cahon 3.9./

.9 ,

4,, REPM LING OPERATIMS

. d'A~i2DgeselcontDrTRATION I.'* ,

.r.9.1

/Frusrds ensaWfGN #0PERWrf0b co .5.fM &$ pron concentrate ini mi =_ aa-4-. of the Reactor Caelant i System,4 .as 1- the___ refueling m m cri canal-/shall u be i.Hmetatained

f m ~w. ---.41

<w ana ui,n 4n.ci-ie q .

4 navi eMc Am,/ djirer4,er as c4e,, 444RJ d .

Eyr a E,,, of U.y or less, e Ad .

A horon concent/tton -;fied of ingnaterpe m e n or equai so sue epism[an -

__- " .*>4 m time Limits "___ m

' APpLICARIL]7Y: , ICDE af,,,wisn sne . ,

r wranc e neas ci ... noirs .-

y. - -- . . _ _ _

EIIM:

Gicnou A) tilth the regul'rements of the above specification not satisfied, immediately

suspend all operations involving CORE ALTERATIONS or positive reactivity-changes and initiat jea iii.. ^ ..iian as gruster w or .y i s as .,

' sien =.i.ining great than er equal to 7000 pr a boren or its equiv until Q is m due to l us than er equal 0.95 or the boma

. c raties is restored to rester than or equal to minimus boron ratiest specified in core Oneratina units art 4te w w <c m A

~ "

. '%on Nrrarc; Lwron doorcro f& 6 wak /<* **fD .

turyttieamerhE0ut29ElY5 .

\ .

.g.1.1 _ I more restrictive of the a e two reactivity consition shan

. rei prior to:

a. Removing or embolting the r vessel head, and u

trithdrenal of any full noth control red in excess f 3 feet f its fully insertad aan inn within the reacter ves 1. -

1

@ f.f././)- 4~ET7h The bores concentration er 1- ==eser s.aoi-t - - - - - t- - -ni.

e -2 e - _.- - ---

~

. _ . _ ._ ;- at leas.t once per 72_ hours.

(L4 D 5.3.1.a su e snais su 1=..new _ _ s.i_ W &_ c_.__. d< - >r_ _s_ w

_. .corre d m,. v. ...

..... ..... ewG..

O c J'""55'd*' #g

/

cA4,e 5 de Tr5 sace ist er per M-less 71 hour8.217593e-4 days 0.0197 hours 1.173942e-4 weeks 2.70155e-5 months st shall be er, verified IIV 131. close W-140.d under administrative control at lessW -176, gy ,.dem .1%2j Bace per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when necessary to askeve to the W durine refuelf stians. f '

'Yorbkelrd Wnler- .

.%ne. TWJuh ,

felvey "

'The r shall be matntataea in whenever fuel is in tne r ctor T

^

vessel th the vessel bead closure ts less than fully tenst orwity the remov yed

. McGUIRE - Ist!T 2 3/4 9-1 Amendment 860. 148

. i

. P, i//

Discussi:n cf Ch:ng:s l S:ctica 3.9 - R:sfusling Operstiens

) ADMINISTRATIVE CHANGES assemblies, therefore, this charge is essentially equivalent and is considered administrative. This change is consistent with

.NUREG-1431.

A.6 The CTS LC0 3.9.8.2 Note that allows the RHR loop to be removed from operation for up to I hour per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during the performance of Core Alterations is being deleted. This allowance whs for initial criticality and is no longer valid. Therefore, this change is considered administrative. This change is consistent with NUREG-1431.

A.7 An exception was added to the CTS 3.9.9 Applicability which allows level in the cavity to be less than 23 feet while latching and unlatching control rod drive shafts. The ITS 3.9.7 exception was mLJe because the definition of Core Alterations includes the unlatching and latching of control rod drive shafts. The control rods are moved slightly during this process, and the definition of Core Alterations includes the movement of any reactivity control component. The control rod drive shafts cannot be latched and

~T unlatched without the level being less than 23 feet. The latching (C and unlatching process could not result in a control rod impacting .

a fuel assembly with sufficient force to cause any significant damage or a fuel handling accident. This change is considered an administrative clarification of existing industry practice. This change is consistent with NUREG-1431.

A.8 - A.10 Not used.

A.11 The CTS 3.9.1 applicability of MODE 6 is modified by a footnote.

The footnote indicates that the reactor is maintained in MODE 6 whenever fuel is in the vessel and the reactor head bolts are less than fully tensioned or the head is removed. The CTS footnote is a repeat of the CTS Table 1.2 footnote definition of MODE 6 and is deleted. The definition of MODE 6 is retained in ITS Table 1.1-1 and ITS Section 1.0. It is unnecessary and inconsistent with the writers guide to repeat the definition of a defined term.

Therefore, the deletion of this redundant information is on administrative format change.

/

A

-()

l McGuire Units 1 and 2 Page A - 23 5/20/97l l !

L_ -_ __ _ !

l 4 -

McGuire and Catawba improved TS Review Comments ITS Section 3.9, Refueling Operations Ca' A ITS 3.9.5, RHR and CC;n Circulation - Low Water Level McGuire' ITS 3.9.6, RHR and Coolant Circulation - Low Water Level M.3.9.6-01 Bases JFD 4 Bases discussion for ITS 3.9.6, Action B.3, page B 3.9-23.

C.3.9.5-01 Bases JFD 4 Bases discussion for ITS 3.9.5, Action B.3, page B 3.9-23.

The Bases discussion for STS 3.9.6 states that the Comp'etion l Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is reasonable based on the low probability of the coolant boiling in that time. This material has not been -

adopted in the corresponding ITS Bases discussion. Bases JFD 4 does not provide a specific explanation for this omission. Comment: Revise the submittal to include this information.

DEC Respones:

DEC can not include this statement because it is not always true. The water sevel in this mode can vary from the vessel flange to just under 23 feet. The time to core boiling is dependent on the volume of water that exists at that time. Administrative controls are in place that limit containment breaches during this time to those that can be closed within the time predicted for core boiling based on the water level. The completion time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is .

,O . this consistent drained).

with water condition since the current level is not licensing routinely maintained basis andlow is (it appropriate for is either being theormajority of opera filled REVISED RESPONSE: l l

The Bases are revised to include the above attemate justification for the completion time as l suggested by the NRC in the comment resolution meeting June 18,1998. l

RHR and Co31 ant Circulation-Low Water Level

B 3.9.6

/ BASES V]

ACTIONS A.1 and A.2 (continued) be OPERABLE and in operation. An immediate Completion Time is necessary for an operator to initiate corrective actions.

L.1 If no RHR loop is in operation, there will be no forced circulation to provide mixing to establish uniform boron concentrations. Reduced boron concentrations cannot occur by the addition of water with a lower boron concentration than that contained in the RCS, because all of the unborated water sources are isolated.

LZ If no RHR loop is in operation, actions shall be initiated innediately, and continued, to restore one RHR loop to operation. Since the unit is in Conditions A and B concurrently, the restoration of two OPERABLE RHR loops and one operating RHR loop should be accomplished expeditiously.

J L3 If no RHR loop is in operation, all containment penetrations providing direct _ access from the containment atmosphere to the outside atmosphere must be closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . With the RHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere. C' losing containment penetrations that are open to the outside atmosphere ensures that dose limits are not exceeded. The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is appropriate for the majority of time during refueling operations, based on time to coolant boiling, since water level is not routinely maintained at. low levels.

SURVEILLANCE SR 3.9.6.1 REQUIREMENTS

. This Surveillance demonstrates that one RHR loop is in

!. operation and circulating reactor coolant. The flow rate is L

determined by the flow rate necessary to provide sufficient

() (continued)

U McGuire Unit 1 B 3.9-23 Supplement 6 l u

RHR and Cool' a nt Circulation'-Low Water Lovel

. B 3.9.6 ff . BASES' SURVEILLANCE SR' 3.9.6.1 (cont'inued) '

REQUIREMENTS decay. heat removal capability,' prevent vortexing in'the.

suction of the RHR pumps, and to prevent thermal and boron stratification in the core. The RCS temperature is determined _to ensure.the appropriate. decay-heat. removal-is-maintained. In addition, during operation of the RHR loop with the' water level in the vicinity of the reactor vessel nozzles, the RHR pump suction requirements must be met. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering the flow.

temperature, pump control, and alarm indications available

' to the operator for monitoring the RHR System in the. control room.

SR '3.9'.6.2.

~

Verification that the required pump is OPERABLE ensures that

an additional RCS or RHR' pump can be placed in operation, if needed, to. maintain decay heat removal and reactor coolant circulation. -Verification is performed by verifying proper breaker alignment and power available to the required' pump.

.The Frequency of 7 days-is considered reasonable in view of other. administrative controls available and has been shown to be acceptable.by operating experience.

REFERENCES .1. UFSAR, Section 5.5.7.

22- 10 CFR 50.36, Technical Specifications,. (c)(2)(ii).

l

l. -.. .

l McGuire Unit 1- B 3.9-24 Supplement 6 c - _ = _ ______-__ _-_

RHR and Coolant Circulation-Low Water Level

B 3.9.6 BASES

[ ACTIONS A.1 and A.2 (continued) t be OPERABLE and in operation. An imediate Completion Time is necessary for an operator to initiate corrective actions.

IL.1 If no RHR loop is in operation, there will be no forced circulation to provide mixing to establish uniform boron concentrations. Reduced boron concentrations cannot occur by the addition of water with a lower boron concentration than that contained in the RCS, because all of the unborated water sources are isolated.

Ib2 If no RHR loop is in operation, actions shall be initiated immediately, and continued, to restore one RHR loop to operation. Since the unit is in Conditions A and B concurrently, the restoration of two OPERABLE RHR loops and one operating RHR loop should be accomplished expeditiously.

LLS If no RHR loop is in operation, all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere must be closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . With the RHR loop requirements not met, the potential exists for the coolant to boil and re. lease radioactive gas to the containment etmosphere. Closing containment penetrations that are open to the outside atmosphere ensures that dose limits are not exceeded. The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is appropriate for the majority of time during refueling operations, based on time to coolant boiling, since water level is not routinely maintained at low levels.

SURVEILLANCE SR 3.9.6.1 REQUIREMENTS This Surveillance demonstrates that one RHR loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient (continued)

McGuire Unit 2 8 3.9-23 Supplement 6 l

RHR and Ccolant Circulatten-Lcw Water Level

B 3.9.6 BASES G(~N SURVEILLANCE SR 3.9.6.1 (continued)

REQUIREMENTS decay heat removal capability, prevent vortexing in the suction of the RHR pumps, and to prevent thermal and boron stratification in the core. The RCS temperature is determined to ensure the appropriate decay heat removal is maintained. In addition, during operation of the RHR loop with the water level in the vicinity of the reactor vessel nozzles, the RHR pump suction requirements must be met. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering the flow, temperature, pump control, and alarm indications available to the operator for monitoring the RHR System in the control room.

SR 3.9.6.2 Verification that the required pump is OPERABLE ensures that an additional RCS or RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the required pump.

The Frequency of 7 days is considered reasonable in view of Other administrative controls available and has been shown

.[R ). to be acceptable by operating experience.

REFERENCES 1. UFSAR, Section 5.5.7.

2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

i. .

O l McGuire Unit 2 8 3.9-24 Supplement 6 E _ _ _

RE and Coolant Circulation-Low Water Level

,' B 3.9.6 BASES O. ACTIONS B.J (continued)

If no RR loop is in operation, there will be no forced circulation to provide mixing to establish uniform boron concentrations. Reduced boron concentrations cannot occur by the addition of water with a lower boron concentration than that contained in the RCS. because all of the unborated water sources are isolated.

L2 If no RHR loop is in operation, actions shall be initiated and continued, to restore one R R loop to immediately,Since the unit is in Conditions A and B operation.

- concurrent 1f. the restoration of two OPERABLE RHR loops and one operating RHR loop should be accomplished expeditiously.

U If no RHR loop is in operation, all containment penetrations providing direct access from the containment atmosphere to-the outside atmosphere must be closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . With the RR loop requirements not met the potential exists for-the coolant to boil and release radioactive gas to the containment atmosphere. Closing containment penetrations-that are open to the outside atmosphere ensures that dose A limits are not exceeded.

t M '

etion h111tvItse or ee b s is reasonable, af the based heilina in that ti .

pvwent uorttva4 W SURVEILLANCE SR 3.9.6.1 p 4 u d..M of- Sc -

REQUIREMENTS This Surveillance demonstrates that one RHR loop is in p mph operation and circulating reactor coolant The flow rate is )

g determined by the flow rate necessary to provide sufficient /

MP pS decay heat, removal capabilityfand to prevent tnemal and wiv , nin.iimion in ma c0rg In addition. during gekren W d b operation of the RHR loop with tM water level in the . -

- vicinity of the reactor vessel noules, the RE pump suction

8#

A//' W g/e M'1

/8 1' requirements must be met. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering the flow, temperature, pump control.

&W)

- h (continued) l WOG STS 8 3.9 23 Rev 1. 04/07/95 (McGarc h 1

,o O

l YL___ _ _ _ _- _ _ -_ __ . _ __. _ .

f INSERT The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is appropriate for the majority of time during refueling operations, based on time to coolant boiling, since water level is not routinely maintained at low levels.

O i

l INSERT Page B 3.9-23 McGuire

__---------,-,__-------------,--,------------,,-----,------,-------------,-------r--

f 4

O ENCLOSURE 5 REVISED ATTACHMENT 3 MISCELLANEOUS CHANGES IDENTIFED BY DEC INTERNAL REVIEW lO i

l

a McGuire & Catawba improved TS Review Comments

~ ITS Section 3.1, Reactivity Control Systems 3.1 Additionalitems 3*I 4 1 ITS Bases 3.1.7, Background and Action C.1.1 and C.1.2 Additional information has been added to the Bases Background and Action C.1.1 and C.1.2 of the DRPI system to clarify system operation with current plant design.

i O

1 f

I mc4_cr_3.1 ' July 7.1998

Rod Position Indication B 3.1.7 4

BASES O l BACKGROUND The axial position of shutdown rods and control rods are 1 (continued) determined by two separate and independent systems: the Bank Demand Position Indication System (connonly called group step counters) and the Digital Rod Position Indication

-(DRPI) System.

The Bank Demand Position Indication System counts the pulses from the Rod Control System that move the rods. There is one's'ep counter for each group of rods. Individual rods in a grobo all receive the same signal to move and should, therefor:, all be at the same position indicated by the group step counter for that group. The Bank Demand Position Indication System is considered highly precise (* 1 step or

% inch). If a rod does not move one step for each demand

' pulse, the step counter will still count the pulse and iricorrectly reflect the position of the rod.

The DRPI System provides a highly' accurate indication of actualz control rod position, but at a lower precision than the step counters. This system is based on inductive analog signals from a series of coils spaced along a hollow tube with a center to center distance of 3.75 inches, which is 6 steps. 'To increase the reliability of the system, the inductive coils are connected alternately to data system A or B. Thus creatin (Data A and Data B)g two separate and independent systems.. Also, th O' reflected six step increments starting at rod bottom.

Because of this arrangement,-the nominal. accuracy of the system is + 3 steps indicated versus true rod position. Due to mechanical positioning of the coils on the rod position detector and expansion in containment atmosphere, another

+ 1 step is added-to system accuracy making it + 4 steps..

If one system fails, the DRPI will go to half accuracy. I f

Data A fails, the accuracy will be + 10, - 4 steps. If Data B fails,-the accuracy will be - 10, + 4 steps. Therefore,-

the maximum deviation between the group demand counters and DRPI could be.10 steps, or 6.25 inches.

APPLICABLE Control and shutdown rod position accuracy is essential SAFETY ANALYSES during power operation. Power peaking, ejected rod worth, or SDM limits may be violated in the event of a Design Basis Accident (Ref. 2), with control or shutdown rods operating outside their limits undetected. Therefore. the acceptance criteria for rod position indication is that rod positions must be known with sufficient accuracy in order to verify the core is operating within the. group. sequence, overlap, design peaking limits, ejected rod worth, and with at least (continued)

!. p McGuire Unit 1 B 3.1-45 Supplement 6 l

Rod Position Indication

^

B 3.1.7 l-l

!e BASES 0 ACTIONS LZ

! (continued) {

Reduction of THERMAL POWER to s 50% RTP puts the core into a condition where rod position is not significantly affecting core peaking factors (Ref. 4). .

The allowed Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is reasonable, based on operating experience, for reducing power to s 50% RTP from full power conditions without challenging plant systems '

and allowing for rod position determination by Required Action A.1 above.

B.1 and B.2 These Required Actions clarify that when one or more rods with inoperable position indicators have been moved in-excess of 24 steps in one direction, since the position was last determined, the Required Actions of A.1 and A.2 are still appropriate but must be initiated )romptly under Required Action B.1 to begin verifying t1at these rods are still properly positioned, relative to their group positions.

If, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the rod positions have not been

(- determined, THERMAL POWER must be reduced to s 50% RTP t,

' within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to avoid undesirable power distributions that could result from continued operation at > 50% RTP, if one or more rods are misa11gned by more than 24 steps. The allowed Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months provides an acceptable period of time to verify the rod positions.

C.1.1 and C.1.2 With one demand position indicator per bank inoperable, the rod positions can be detennined by the DRPI System. Since normal power operation does not require excessive movement of rods, verification by administrative means that the rod position indicators are OPERABLE and the most withdrawn rod and the least withdrawn rod are s 12 ste)s apart within the 1 allowed Completion Time of once every 81ours is adequate. .

Since DRPI is the only operable rod position indication, administrative means are actions taken by the control room SR0 to assure that the DRPI for the affected bank remains operable at all times. These administrative means would prevent any maintenance or testing of the operable DRPI for the affected bank until the inoperable demand position indicator is returned to operable status.

t n n (continued)

'd l McGuire Unit 1 B 3.1-48 Supplement 6 I

---__ _ - - _ - - - - _ - - - _ . - -- -_ - - - - - - - - - - - - - - - - - - - - -- - - - - - - - ------------J

Rod Position Indication B 3.1.7 4

.n BASES l

BACKGROUND The axial position of shutdown rods and control rods are (continued) determined by two separate and independent systems: the t Bank Demand Position Indication System (commonly called group)

(DRPI System. step counters) and the Digital Rod Position Indication The Bank Demand Position Indication System counts the pulses from the Red Centrol System that move the rods. There is one step counter for each group of rods. Individual rods in a group all receive the same signal to move and should, therefore, all be at the same position indicated by the group st?o counter for that group. The Bank Demand Position Indication System is considered highly precise (* 1 step or i % inch). If a rod does not move one step for each demand pulse, the step counter will still. count the pulse and incorrectly reflect the position of the rod. .

The DRPI System provides a highly accurate indication of actual control rod position, but at a lower precision than the step counters. This system is based on inductive analog signals from a series of coils spaced along a hollow tube with a center to center distance of 3.75 inches, which is 6 steps. To increase the reliability of the system, the inductive coils are connected alternately to data system A or B. Thus creatin (Data A and Data B)g two separate and independent systemsAlso, the.

reflected six step increments starting at rod bottom.

Because of this arrangement, the nominal accuracy of the system is + 3 steps indicated versus true rod position. Due to mechanical positioning of the coils on the rod position detector and expansion in containment atmosphere, another

+ 1 step is added to system 3ccuracy making it + 4 steps.

If one system fails, the DRPI will go to half accuracy. If Data A fails, the accuracy will be + 10. - 4 steps. If Data B fails, the accuracy will be - 10, + 4 steps. Therefore, the maximum deviation between the group demand counters and DRPI could be 10 steps, or 6.25 inches.

APPLICABLE Control and shutdown rod position accuracy is essential SAFETY ANALYSES during power operation. Power peaking, ejected rod worth, or SDM limits may be violated in the event of a Design Basis Accident (Ref. 2), with control or shutdown rods operating outside their limits undetected. Therefore, the acceptance criteria for rod position indication is that rod positions must be known with sufficient accuracy in order to verify the core is operating within the group sequence, overlap, design peaking limits, ejected rod worth, and with at least

! l l

(continued)

McGuire Unit 2 B 3.1-45 Supplement 6 l l

NL__ _ . _______ _ _ _ _ _ _ _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

Rod Position Indication B 3.1.7 e

a BASES ACTIONS L2 (continued)

Reduction of THERMAL POWER to s 50% RTP puts the core into a condition where rod position is not significantly affecting core peaking factors (Ref. 4).

The allowed Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is reasonable, based on operating experier,ce, for reducing power to s 50% RTP from full power concitions without challenging plant systems and allowing for rod position determination by Required Action A.1 above.

8.1 and B.2 These Required Actions clarify that when one or more rods with inoperable position indicators have been moved in excess of 24 steps in one direction, since the position was last determined, the Required Actions of A.1 and A.2 are still appropriate but must be initiated promptly under Required Action B.1 to begin verifying that these rods are still proper'ly positioned, relative to their group positions.

If, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the rod positions have not been p)

-C determined, THERMAL POWER must be reduced to s 50% RTP within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to avoid undesirable power distributions that

-could result from continued operation at > 50% RTP, if one or more rods are misaligned by more than 24 steps. The allowed Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months provides an accep, table period of time to verify the rod positions.

C.1.1 and C.1.2 With one demand position indicator per bank inoperable, the rod positions can be determined by the DRPI System. Since normal power operation does not require excessive movement of rods, verification by administrative means that the rod position indicators are OPERABLE and the most withdrawn rod and the least withdrawn rod are s 12 steps apart within the allowed Completion Time of once every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is adequate.

Since DRPI is the only operable rod position indication, administrative means are actions taken by the control room SRO to assure thac the DRPI for the affected bank remains operable at all times. These administrative means would prevent any maintenance or testing of the operable DRPI for the affected bank until the inoperable demand position indicator is returned to operable status.

i (continued) l B 3.1-48 Supplement 6 l McGuire Unit 2

l

' Rod Position Indication 6

B 3.1 BASES s

BACKGROUND The axial position of shutdown rods and control rods are (continued) determined by two separate and independent systems: the Bank Demand Position Indication System (commonly called roup step counters) and the Digital Rod Position ndication (DRPI) System. g '

The Bank Demand Position Indication System counts the pulses from the Rod Control System that move the rods. There is one step counter for each group of rods. Individual rods in a group all receive the same signal to move and should, therefore. all be at the same position indicated by the group step counter for that group. The Bank Demand Position Indication System 1s considered highly precise (* 1 step or i % inch). If a rod does not move one step for each demand pulse, the step counter will still count the pulse and '

incorrectly reflect the position of the rod.

The DRPI System provides a highly accurate indication of actual control rod position, but at a lower precision than the step counters. This system is based on inductive analog signals from a series of coils spaced along a hollow tube with a center to center distance of 3.75 inches, which is 6 steps. To increase the reliability of the system. the inductive coils ara c^anactad altarnatalv tn riata svst =n A or B. ThusfTf o ystea fails, the will go on wir' r accuracy wit effective coil ng of 7.5 inche which is 12 st . Therefore, the 1 indication ac acy of the System is i 6 st (* 3.75 inches), a the mum uncertainty 1 jj k 2 steps (t 7.5 in . With an 7 ff)gg indicated deviati 12 steps between t roup step counter and DRP he maximum deviation ween actual position a demand posittor. cou . 24 steps, or 15 i f 1 APPLICABLE Control and shutdown rod position accuracy is essential '

SAFETY ANALYSES during power operation. Power peaking, ejected rod worth, or SDH limits may be violated in'the event of a Design Basis Accident (Ref. 2), with control or shutdown rods operating outside their limits undetected. Therefore the acceptance criteria for rod position indication is that rod positions must be known with sufficient accuracy in order to verify 2 the core is rating within the group sequence. overlan_

design peaki limits, ejected rod worth, and withAmini d SDH (LCO 3.1

Shutdown Bank Insertion Limits." and (continued) g & STS-- B 3.1 47 Rev 1. 04/07/95 hct M d

Insert i i

Thus, creating two separate and independent systems (Data A and Data B).

Also, the coils are not placed at the reflected six step increments O starting at rod bottom. Because of this arrangement, the nominal

accuracy of.the system is + 3 steps indicated versus true rod position.

Due to mechanical positioning of the coils on the rod position detector and expansion in containment atmosphere, another

~ + 1 step is added to system accuracy making it + 4 steps.

If one system fails, the DRPI will go to half accuracy. If Data A fails, the accuracy will be + 10. - 4 steps. If Data B fails, the accuracy will be .10, + 4 steps. Therefore, the maximum deviation between.the group demand counters and DRPI could be 10 steps, or 6.25 inches.

O i

4 I

v'

,y O

Insert 8 3.2-47 1

i l Rod Position Indication ACTIONS M (continued)

Reduction of THERMAL POWER to s 50t RTP puts the core into a condition where rod position is significantly affecting core peaking factors (Ref.

, The allowed Completion T1 f 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is reasonable, based I

on operating experience, for reducing power to s 50% RTP from full power conditions without challenging plant systems and allowing for rod position determination by Required Action A.1 above.

B.1 and B.2 i

These Required Actions clarify that when one or more rods with inoperable position indicators have been moved in excess of 24 steps in one direction, since the position was last determined, the Required Actions of A.1 and A.2 are still appropriate but must be initiated promptly under Required Action B.1 to begin verifying that these rods are still properly positioned, relative to their group positions.

If, withik4 hours. the rod positions have not been determined. TERMAL POWER aust be reduced to s 50% RTP within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to avoid undesirable power distributions that could result from continued operation at > 50% RTP, if one l O or more rods are misa11gned allowed Completion Time ofsy more than 24 steps. The rs provides an acceptable j period of time to verify the rod positions.

C.1.1 and C.1.2 With one demand position indicator per bank inoperable, the rod positions can be determined by the DRPI System. Since -

normal power operation does not require excessive movement of rods, verification by administrative means that the rod position indicators are OPERABLE and the most withdrawn rod and the least withdrawn rod are s 12 s s apart within the rs is adequate. '

Instithallowed Completion Time of once every 8 ly (continued) 4dQG-STS'~ B 3.1 50 Rev 1. 04/07/95 hcM<.,

Insert O Since DRPI is the only operable rod position indication, administrative means are ' actions taken by the control room SR0 to assure that the DRPI for the affected bank remains operable at c11 times. These administrative means would prevent any maintenance or testing of the operable DRPI for the affected bank until the inoperable demand position indicator is returned to operable status.

i e

O i

O re ert e 3.2-eo I

c.-_. ._. _ _ _ ._ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _____

v t e Letter Sequence Other
TAC:M98964, Implement 10 CFR 50, Appendix J, Option B, TSTF-45, TSTF-17 (Open) TAC:M98965, Implement 10 CFR 50, Appendix J, Option B, TSTF-45, TSTF-17 (Open)
Initiation Acceptance... Administration Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting Questions 30 January 1998 28 April 1998 Answers Results Approval... Other: ML20151Y172, ML20151Y235, ML20195K191, ML20199E417, ML20199J107, ML20204E179, ML20205B616, ML20205M674, ML20216B753, ML20216B957, ML20217B718, ML20217E628, ML20217E973, ML20236M023, ML20236M874, ML20248L685, ML20248L705, ML20249C577
MONTHYEAR1998-06-25 [Table View]

ML20236M874

Text

Subject:

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