Text

Proposed Tech Specs Sections 3.1,3.2,3.4,3.5 & 5.0 Converting to ITS
	ML20155C070
Person / Time
Site:	Diablo Canyon
Issue date:	10/23/1998
From:	; PACIFIC GAS & ELECTRIC CO.
To:	;
Shared Package
ML16342A557	List: ML20155C070;
References
	NUDOCS 9811020070
	Download: ML20155C070 (40)
	v • d • e

1 1

1 REACTIVITY CONTROL SYSTEMS g PCSITION INDICATION SYSTEMS - OPERATING e T U LIMITING CONDITION FOR OPERATION 3.1.3.2 The Digital Rod Position Indication I IndicationSystems%11beOPERABLEandcapab{.ystemandtheDemandPositign Of deter ~ "4ng the cc",tr0 rod 13 01.Le

" 0 5 ' t : 0". : 'M'" ; stops.

APPLICABILITY: MODES 1* and 2* 33 .og.Ls20 l ACTION:

a. With a maximum of one digital. rod position indicator per bs4 group inoperable for one or more: groups either: 13-02-LS15 l

1. Determine the position of the nonindicating rod (s) indirectly by the movable incore d 13-03-LS12 onceperBhkursand'T0diatoyw?tectorsatleast ithin 4-hours af r any motion o the nonindicatin6 rod which exceeds steps in one direction since tfie last determination of the rod's position, or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL 13-04 #

POWER within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in-Hot Standby-within t1e next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. W' t1 more than:one^'didital" rod'p65itionTindicatoFper group in6perable

" ~ ' ' '~

e tier: ' ' ~ '

'1'.arDetermihe"the posi ion "ofitheMonindicating r'ods 13-08-Ls20 indirectly by the movab e incore: detectors at least once (q< and with n hoursn.after'any min otion V ber8 hour rod wh onindicat exceeds 24 stepp onedire of t ~ n a nce the-.7 {t determynation;ofithe; rod s;positiont @.g Q pg%w

'h t 0t digita rod; position:. indicator::pergroup+;1nino.perable,or um 0" one 13-08-LS20

, M *ecy r *

NcSo m% =. - ~- -

t D" J ~ 2."BellinJH0Tl$TANDBY Within!the;nekt;6; hours? 13-08-Ls20 b c. With a maximum of one demand position indicator per bank inoperable either:

1. Verify that all digital rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 12 steps of each other at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . or

2. Re uce THERMAL POWER to less tlan 50% of RA"ED THERMAL 13-04-M P0 ER within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in iot. Standby w1 thin tie next

6. ours. 9811020070 981023 SURVEILLANCE REQUIREMENTS PDR ADOCK 05000275 P PDR 4.1.3.2 Each diq1tal rod Dosition indicator shall be determined to be OPERABLE by verifyingthattheDemandPositionIndicatignSystemandthgDigitalRodPosition b lcation Syste g grg wit,h g 12,stegs g ; g ggg^,pp" l y g g gxc g d gr'"j ,, 3 3 07.y

G'nA4;;i BA; WJ;m Y W ;4 ; ; M ; ' 6 % " ;; # ; L ;~'n i; W ; r o A ; n ; ; m ; ; 'T;, 4 ; W ,; ; ' ~ "

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Separate condition entry is allowe"d~for each inoperable rod position indicator and each demand. position indicator per bank. 13-08-LS20 4.l. L 3 Inw c M Ws DIABLO CANYON - UNITS 1 & 2 3/4 1-18 TAB 8.4A

DESCRIPTION OF CHANGES TO TS SECTION 3/4.1 (Continued)

! CHANGE NUMBER NSH.C DESCRIPTION 13-06 A Not applicable to DCPP, See Conversion Comparison Table.

(Enclosure 3B) 13 07 M The proposed modifications to the SR would require a venfication of agreement between digital and demand indicator systems prior to criticality after each removal of the reac.'or vessel head, instead of every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This reflects a reorganization vf SRs in the ITS. The requirement for a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months comparison would be moved to SR 3.1.4.1 in the ITS. The post-vessel head removal requirement would be a new specification that demonstrates rod position system OPERABILITY based on a comparison of indicating systems throughout the full range of rod travel. The Frequency requirement of prior to criticality after each !

removal of the reactor vessel head would permit this comparison to be i performed only during plant outages that involve plant evolutions (vessel head removal) that could affect the OPERABILITY of the rod position indication systems. The Frequency change is based on Traveler TSTF-89.

13-08 LS20 Adds provision from Callaway's current specifications which would, under certain conditions, allow continued operation with more than one inoperable DRPI per grcup. A separate Condition entry allowance is permitted for each inoperable rod position indicator per aroup and each sed Traveler T -23

demand

"'c

.. i-if-- position indicator O=---mar ^^'OC) perL, bank. (A prp"rf 1 is in saina t _pr l (covdis issue.J -2o k hd 13-09 LS23 Not applicable to DCPP. See Conversion Comparison Table

(Enclosure 38).

14-01 R phefhutdown Pfsition indieption System Spbcification 3.1/3 ils yel#cated outside of the TS/This is consistient with NURES-1431.)

15 01 [ The Rod The RCS mperature p Time S cation 3.1.3.4 is re and reactor coolan umps operating ed outside of the 0 require t for rod p testing are combin with CTS Surveill ce 4.1.3 , then inco rated into ITS SR 3.1 This is consist with ;

NU EG-1431.J__p. Not used '

15-02 A The R Drop Time 4.1.3.4.a is to the Contro odITS O

.1. SR 3.1.4. is chang @ tent with N -143

h apphewe k Ocpp, s;<cearsec,4,.re%q.fmwar n),fcs.i. ,_

16-01 LS14 This TS would be revised to apply to shutdown " banks" instead o shutdown " rods;" this is consistent with NUREG 1431. The current ACTION statement permits one rod to be inserted beyond the limits; the proposed ITS Condition A would allow one or more banks to be inserted beyond the limit.

DCPP Description of Changes to Current TS 10

O CONVERSION COMPARISON TABLE - CURRENT TS 314.1 O O Page 8 of 10 TECH SPECH CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO COMANCHE WOLF CREEK CALLAWAY CANYON PEAK i

13-04 A requirement would be added to bring the plant to Yes Yes Yes Yes M MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months if the required ACTIONS and Completron Tirnes are not met.

13-05 The proposed change would retain an ACTION No, not in CTS - No, not in CTS - Yes Yes A statement, currently in the plant TS, that permits see 13-08-LS20. see 13-08-LS20.

continued POWER OPERATION with more than 1 digital rod position indicator per group inoperable, 13-06 The change would allow separate Condition entry for No, not in CTS - No, not in CTS - Yes Yes A each inoperable DRPI per group or each demand see 13-08-LS20. see 13-08-LS20.

indicator per bank.

13-07 The proposed modifications to the SR would verify Yes Yes Yes Yes M agreement between digital and demand indicator systems prior to criticality after the reactor vessel head was removed instead of every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The Frequency change is based on Traveier TSTF-89.

13-08 Adds provision in Callaway's current specifications which Yes Yes No, already in No, already in LS20 would, under certain Conditions, allow continued CTS. CTS.

operation with racre than one inonerable DRPI Der QrouD. __

phb b cc=b:x: .-r ' r;.;:;;"!M ??, "r *~*y - 2:1)) 1 /5 06 /

13-09 CTS ACTIONS b.1.b) and b.1.c) of LCO 3.1.3.2 are No, not in CTS. No, not in CTS. Yes Yes LS23 deleted. SDM is ensured in MODES 1 and 2 by rod position. Multiple inoperable DRPIs will have no impact on SDM in MODES 1 and 2 if the control rod positions are verified by attemate means and rod motion is limited consistent with the accident analyses. Deletion o."these =

requirements is consistent with traveler @^C 7^,, Rd.' *2 R 3d' #

14-01 Relocates CTS 3.1.3.3 to licensee controlled documents, es ee 95- Yes, relocated to No, see No, see R consistent with NUREG-1431. 07 ated /4/95, TRM. Amendment 89. Amendment 103.

L 95 22.

Y No, se Aw wwis 8 1- 21 i20/na DCPP Conversion Comparison Table - Current TS

indu:try Travelers Applicabb t3 Section 3.1 O

G TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF-9, Rev.1 incorporated 3.1-1 NRC approved.

TSTF 12, Rev.1 incorporated 3.1-15 NRC approved. ITS Special Test Exception 3.1.10 is retained and re-

} numbered as 3.8.1, consistent with this traveler and TSTF-136 TSTF-13, Rev.1 incorporated 3.1-4 NRC approved.

TSTF-14, Rev.@4 incorporated 3.1-13 NRC approved )

TSTF 15, Rev.1 Incorporated N/A NRC approved. l

1 TSTF-89 incorporated 3.1-8 NRC approved.

TSTF-107 Re.v. )

3 incorporated 3.1-6 TSTF-108, Rev.1 hcorporated @ 3. /- R/

@NRC appro disi"iiW3allb 72I/-col l TSTF-110, Rev. $1. Incorporated 3.1-10 Nac. 4;pm/ed @375]

TSTF-136 Incorporated 3.1-9,3.1-15 TSTF-141 Not Incorporated N/A Disagree with change;

,[m '

traveler issued after cut-

\ off date m TSTF-142 @Tcorporated @ Tim;;;-iredbiIY.D 3.I-27. eff4ete.NRC @W.

k'QW73, % 17tTc2EQ Jr(corporated/ 3,Vf _/ / Gb5.I.h WOG-105 Incorporated 3.1-16 h i

e

?

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i v

DCPP Mark-up of NUREG 1431, Rev.1

Rod Position Indication 33 ~9 3-M 3]1;7 3.1 REACTIVITY CONTROL SYSTEMS 34-8 3.112 Rod Position Indication LC0 3- M 3^.'1;7 "

?The Digital Rod Position Indication (DRPI) System and the

'" Demand Position Indication System shall be OPERABLE. g.Ps APPLICABILITY: MODES 1 and 2.

1 ACTIONS 1 iT5

.......................N0TE---------------------~~----------------------

Se arate Condition entry is allowed for each inoperable rod position

, ^b in icator per group and each demand position indicator pes. bad . (t y7 .

3, CONDITION REQUIRED ACTION COMPLETION TIME A. One E04RPI per group A.1 Verify the position of the Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months inoperable for one or rods with inoperable B-PS more groups, lpositi.on indicatorsindirectly by using movable 3.1 12 incore' detectors.

@ 8 hcurs 4

A.2 Reduce THERMAL POWER to s 50% RTP. l l

B More than~one DRPIfp~#r fM Verify 7ths?b6sitiod?bfiths Onbe:p^

~ " ' ~ " 'ERB!houd

' " 3'1 7 group'imperablef^

~

T rodstwithsinoperable ~~~

@ position sindicators^ id7able indirectiylbykusin incoreldetectors7"gii allQ 24! hoers B : Rest 6rdHn~o~~

Eindicators peNbisto 6siti6n to:0PERABLE'~~^

~ ~ ' ~ '

3'1'12 statusisuch thatta! maximin oftoneORPljpeggroupjis jnoperablej B C. One or more rods with BC.1 Verify the position of the 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 3.1 17 inoperable ^0a tion rods with inoperable B

!r.mcatersBRPIshave position indicators been moved in excess of indirectly by usin9 3.1-12 24 steps in one movable incore detectors. I direction since the last determination of M the rod's position.

BC.2 Reduce THERMAL POWER to 8 hcurs s 50% RTP.

8' f1GC9 4ht.Centrol arcis ureer g'~

mCM~l cehtI. WmatamY 61!E o3i r.c ,

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62 Ak N IDr' ca d (t & d. Oyc p I hour rece w cce n Af mm 2 3 ;

u DCPP Mark-up of NUREG-1431. Rev. 1 3.1-14

1 Rod Position Indication l B 3-1-8 3.L7' ~~

BASES !

i

' ACTIONS .A 1 *

(continued)

When one DRPI per gro p fails, the position of the rod may still be determined indirectl y l Required.- Action :may al .y o .be useensuringof.the at..leastionce movable.incore Der-hours detectors.

that fo The i

- satisfies LCO 3.2.1; F ' satisfies'.LC0:3 2.2; and SDH is within the limits provided'in thekOLR 'provided the;nonindicating rodsihave.not l been moved." Based on experience 7 normal ~ power ooeration does not require excessive movement of banks. If a bank has been si moved. the Required Action of C.1 or C.2 below is required.gnificantly Therefore. verification of RCCA" position within the Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is adequate for allowing continued full power operation, since the probability of simultaneously having a rod significantiv out of position and an event sensitive to that rod position is small.

hd 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. 3).

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 d Bl2fG.h daA) f% e M WhenTmor6Tthan~one:0RPIT

\ necessaryi:to ensure that:peFgFoupTfillB^additfonallactions acceptables powern distribution limits: are are.

ma i ntained a mi nimum SDM ;i s maintained nand >lyser a re rod misalignmentxonTassociatediaccidentiana e.Mm1 ted indirect position: determination:availabletvia: movable (1ncore detectorp ,

wi11 minim 1.ze..thejpotentialffortrodyisalignment;

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N (Tnse~+ ) f .

'picg - +e. Red. Cm+n>l %skm in manud}

$$Sures unplcnnat. rod rm%n will nof OLCu'. ]

(Continued) l l

l l

O DCPP Mark-up of NUREG-1431. Rev.1 Bases B 3.1-30

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

insert for revised FLOG Response Q3.1-20 ITS Section 3.1 - Enclosure SB - page 83.1-30 The immediate Completion Time for placing the Rod Control System in manual reflects the urgency with which unplanned rod motion must be prevented while in this Condition.

Monitoring and recording reactor coolant T,y, help assure that significant changes in power distribution and SDM are avoided. The once per hour Completion Time is acceptable because only minor fluctuations in RCS temperature are expected at steady

. state plant operating conditions.

1 f

1 i

l

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

i l

JUSTIFICATION FOR DIFFERENCES FROM NUREG-1431 NUREG-1431 Section 3/4.1 This Enclosure contains a brief discusssordustification for each marked-up technical change to NUREG-1431, to make them plant specific or to incorporate generic changes resulting from the Industry /NRC genenc change process. The change numbers are referenced directly from the NUREG-1431 mark-ups (Enclosure SA). For Enclosures 3A, 3B,4 GA, and 6B text in brackets "[ ]* indicates the information is plant specific and is not common to all the JLS plants. Empty brackets indicate that other JLS plants may have plant specific information in that location.

CHANGE NUMBER JUSTIFICATION 3.1-1 in accordance with TSTF-9, Rev.1, this change would relocate the specified limit for SDM from ITS to the COLR. This change occurs in several specifications including the ,

specification for SDM and those specifications with ACTIONS that require verifyin SDM within limits.

3.1-2 IThe Note for S .1.2.1 indicates t predicted rea ty values mm be adjus (normalized) correspond to th {

asured core re 'vity prior to exceedin fuel bumup of FPD after each fueling. However th the Bases for S ication 3.1.3 an CTS requireme in Specification .1.1.5 state that the no i

lization shall i

! be do pnor to exceeding fuel bumup of 60 PD after each refuelin )

Noe Used, ,

3.1-3 % :;7 '" t OOPP 'M Or f: CrF;t. T i %. :'::= SC;%

1 3.1-4 SR 3.1.4.2 of NUREG-1431, Rev.1 would be deleted. In accordance with TSTF-1 the intent of this SR is only to determine the next frequency for SR 3.1.4.3. 3.1 -oo C. l e ormance of SR 3.1.4.2 is not necessary to assure that the LCO is met; SR 3.1. --

fulfills that purpose. Therefore, SR 3.1.4.2 may be deleted. In addition, the note in the frequency column of SR 3.1.4.2 would be moved to become Note 1 in the surveillance column of SR 3.1.4.3. This is for clarification purposes. As discussed in CN 3.1-9, section renumbering results in SR 3.1.4.3 of NUREG-1431, Rev.1 becoming ITS SR 3.1.3.2.

l 3.1-5 Per CTS [3.1.3.1], the words *with all* have been removed from ITS LCO 3.1.4. This is a clanfication that ensures the proper interpretation of the LCO. The change makes it clear that only one channel of DRPI is necessary to meet the alignment accuracy requirement of the LCO. With the word "all" in the statement, it may be possible for those unfamiliar with the DRPI design to interpret the LCO as applying to all channels of DRPl.

3.1-6 LCO 3.1.4 would be split into two separate statements to clarify that the alignment limit is separate from OPERABILITY of the control rod. The Condition A wording is broadened from *untrippable" to " inoperable

to ensure the Condition encompasses all causes of inoperability. Previous wording was ambiguous for rods that, for instance, had slow drop times but were still tnppable. These slow rods are inoperable rods, and the change clarifies the appropriate ACTIONS. The Bases are changed to reflect the changes to the LCO and Condition A. These changes are based on TSTF-107, 3.1-7 This change to the ISTS would incorporate, into LCO 3.1,7, an ACTION statement that was previously approved as part of the Callaway and Wolf Creek licensing basis @

5"!. C t- -

2. The ACTION statementwould permit continued POWER

- OPERATION for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months with more than one DRPl channel per rod group @ 20 inoperable. The ACTION statement specifies additional Required ACTIONS beyond l those applicable to the Condition of 1 DRPI channel per group inoperable. The Bases for this chance also would be incorpci.;.d into tr e Bases for the plant ITS. /These c es are nsistent velerc--JJ n 7 The note er the ACTION is anged consisten equipns. t i O theneWw "O DCPP Description of Changes to improved TS 1

CONVERSION COMPARISON TABLE FOR DIFr cRENCES FROM NUREG-1431, SECTION 3/4.1 P ,I1 of 3 .

t TECHNICAL SPECIFICATION CHANGE APPLICABILITY l NUMBER DESCRIPTION DIABLO CANYON COMANCHE WOLF CREEK CALLAWAY t PEAK 4

3.1-1 In accordance with industry Traveler TSTF-9, Rev.1,this Yes Yes Yes Yes change would relocate the specified limits for SDM from ,

j severalTS to the COLR. t 3.1-2' [ Changes note to SR 3. 'Z.1, which deals th Y Vee- Ves No-maintaimng 9 ,- i f _ . .g .

verifying e reactivity in limits, to state at the uA NA lTO n . 4 lTO s ...g. f nor tion of pr ~ reactivity values corresporu l NA NA __ i i i to red values 11 be done prior to xceeding a gy, j,q '

( up of 60 EF after each r MhofU5PM *

! 3.1-3 The Creek ITjVLCO 3.1.6 Reqpired Action .1is 44e- No- Vee- -Ner l revi from "Be in MODE 3." to *p6 in MODE with ly NA N4 MA NA S 7./- 2 3 t l

j i

~ f-+ Na+ used ;

j. 3.1-4 in accordance with industry Traveler TSTF-13/1G(T) Yes Yes Yes Yes !

' SR 3.1.4.2, which requires venfying MTC within the 300 3 ppm boron limit, is deleted and the note in that SR is TEll-ab ) !

4 moved to the SR that requires the lower MTC limit to be :

verified. The deleted SR is not a requirement separate l

. from the lower MTC venfication SR, but is essentially a I

clarification of when the SR for the lower MTC limit should - l i

be performed

! 3.1-5 Per CTS [3.1.3.1] the words "with all" are removed from Yes Yes Yes Yes l the LCO for control rod alignment limits. This ensures that the number of channels of DRPl required to be l r

OPERABLE will not be misconstrued i

in accordance with TSTF - 107, the change provides Yes Yes Yes Yes

i. 3.1-6 ,

i additional clanfication that the alignment limits in the LCO l are separate from the OPERABILITY of a control rod.

t

! 3.1-7 An ACTION statement that was previously approved as Yes Yes Yes Yes !

part of the current licensing basis of Callaway and Wolf r Creek would be added to ITS 3.1.7(* r:rr? @ q aa-m I O.,,.x a_7. The ACTION statement would permit operation for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months with more than one digital ,

rod position indicator per group inoperable.

I i

i i DCPP Conversion Comparison Table -Improved TS j

________ . _i

Enclosure 2 l PG&E Letter DCL-98-154 ;

ADDITIONAL INFORMATION COVER SHEET )

l ADDITIONAL INFORMATION NO: O 3.2-3 APPLICABILITY: CA, WC, DC, CP REQUEST: ITS 3.2.1 Heat Flux Hot Channel Factor l

CTS 3/4.2.2 Heat Flux Hot Channel Factor (All FLOG Plants)

DOC 02-06-A JFD 3.2-12 ITS SR 3.2.1.1 & 3.2.1.2 Frequency Comment: The ITS SR frequency has been changed from the STS frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This is based upon the incorrect justification that the CTS would :

allow 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months based upon ITS SR 3.0.3, since the CTS does not specify a frequency. l Adopt the STS SR frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

FLOG RESPONSE (original): The change descriptions (DOC 2-06-A & JFD 3.2-12) will be revised to provide a basis for the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months that is predicated on the time required to perform the surveillance.

Callaway and Wolf Creek are incorporating this change (DOC 02-06-A, JFD 3.2-12) in lieu of maintaining CTS which did not specify any completion time. DOC 02-13-LG (applicable to Callaway only) and JFD 3.2-17 are no longer used, n FLOG RESPONSE (supplement): As discussed in a telecon with the NRC staff on October 1, Q 1998, additional justification for the basis of the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months surveillance frequency has been added to JFD 3.2-12.

l Additionally, this item is related to Comment Number O 3.2-7 for Callaway and Wolf Creek. No additional response is required for Comment Number Q 3.2 7. l ATTACHED PAGES:

Attachment 8 - CTS 3/4.2 / iTS 3.2 l

Encl. 6A 2 l

i JUSTIFICATION FOR DIFFERENCES FROM NUREG 1431 NUREG 1431 Section 3/4.2 3.2-08 Consistent with Traveler TSTF-99, the LCO 3.2.1 (Fo Methodology), Required ACTION B.1.

Completion Time for the reduction of the AFD limits if F*o(Z)is not within limits is increased from 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This makes it consistent with the Completbn Time associated with Required ACTION A.2. of LCO 3.2.1 (Fxy methodology). The change is acceptable because it eliminates an inconsistency in the ITS.

3 2-09 For consistency with CTS 3.2.4 and ITS 3.3.1, Condition D, the breakpoints for the Applicability of the surveillances in the notes in ITS SR 3.2.4.1 and SR 3.2.4.2 are modified to be applicable at less than or equal to 75 percent RTP, and greater than 75 percent RTP, respectively. T '

administrative change that retains CTS requirements 2-4 ard is con 6ititni & TSTF-24),

3.2-10 Consistent with Traveler TSTF-110, this change moves requirements for increased surveillance l

frequencies in the event of inoperable alarms to licensee controlled documents. This change is i acceptable because it removes requirementh regarding alarms and alarm responses that are not necessary to be in the TS to protect public health and safety. gg 3.2-11 Not apphcable to DC See Conversion Comparison Table (Enc'esure 68).

T 3.2. .).2., bed en p\cd (getienct3 l 3.2 12 --Cone,;terd ..? CTS, 'he required time for completion a flux ma for determination of the j heat flux hot channel factor is changed from 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to hours achieving equilibrium Conditions. The proposed change affects SR 3.2.1.1 and S R he proposed time (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) is a reasonable time period forithe compt on of the survetI and does not ow for I plant o on in an unce n condition for rotracted time peri This change is fl consis t with the TS r uirements of S fication 3.0.4 (and ssociated Bases) t t allow I p 24 urs for the com tion of a surveill ce after prerequisji er plant conditions ar attained ,

t -

d for which an ception to Specif tion 4.0.4 was provjdedJ InsecU 3.2-13 This change retains the CTS for the performance of peaking factor determinations following plant shutdowns. The CTS, through the exemption to Specification 4.0.4, allows prerequisite plant conditions to be obtained prior to requiring that the surveillance be completed.%

3.2 14 Not applicable to DCPP. See Conversion Comparison Table (Enclosure 68). hd S 3.2-15 This change incorporates Traveler TSTF-109. ACTION A.2. would require the OPT determined rather than performing a specific surveillance because more than one surveillance can be used to determine OPTR. SR 3.2.4.1 was revised to retain allowance that SR 3.2.4.2 may be performed in lieu of SR 3.2.4.1. -The ncic for SR 12A2 M chWed ! :#e -

pe"c= e ? cm "er ere" Om tt: c ',nspcci ..? " ' 75 ps;:n; These !

changes are acceptable because they clarify the ITS regarding frequency and us of incore l flux monitoring for QPTR measurement. The changes reflect that incore detectors rovide an acceptable OPTR determination during all plant Conditions. '

3.2-16 This change would require both transient and static Fo measurements be determined when performed for Required ACTIONS 3.2.4 A.3 ano A.6. The intent of the Required ACTIONS is to verify that Fo(Z) is within its limit. Fo(Z)is approximated by F8(Z)(which is obtained via S 3.2.1.1) and FE(Z) (which is obtained via SR 3.2.1.2). Thus, both F8(Z) and F*o(Z) must be established to venfy Fo(Z). This change is consistent with Traveler WOG 105.

3.2 17 LNot appbcableio DCPP Se(Conversion Corfipanson Tably(Enclosp/e 68)] 03.2-3 h utu -

3.2-18 Not appicable to DCPP. See Conversion Comparison Table (Enclosure 6B).

Q 3.2 19 Not appicable to 0CPP. See Conversion Comoarison Table (Endosure 8m h rett ord becpeg3 4ec % S.2. A2. me revised corsitJtnt Wn ~

hesccd Cre/=entoke bremds h+ preside. be a period et time, cer4ec r 0 DCPP Description ofMtcuoM Cuang ~es ccedA,.WnS f m.pce.c. v 3.1- 2. 0 No1 appnic&te %ccep, see CowerSon Comparison TcMe (Endmre %.

tQ

1 l

insert for Supplemental FLOG Response Q3.2-3 ,

ITS Section 3.2 - Enclosure 6A - page 2 Insert a for JFD 3.2-12:

A flux map is taken after a power level increase greater than a specified amount to verify l Fo is within limits and to provide assurance that Fo will remain within limits until the next '

required flux map is taken. Based on plant experience, the flux maps taken during power ascension provide a high degree of confidence that Fo will be within limits at the next power plateau. As such, the exact time period allowed for performance of the surveillance, after reaching equilibrium, is not a significant safety consideration. The proposed time (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months )is a reasonable time period for obtaining and evaluating a flux map and then completing the procedural steps associated with this surveillance.

Further, the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months time period provides a reasonable limit on the length of time that the plant can operate in an unconfirmed condition, i

1 i

i

Insert for Q3.2-3

! Enclosure 6a - page 2 INSERT for 3.2-12: i

! I obtaining and evaluating a flux map and then completing the procedural steps associated with this i surveillance. Further, the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months time period does not a: low for plant operation in an uncertain !

l condition for a protracted time period. !

l'

s.

l I

l 1

1 0

1

Enclosure 2 PG&E Letter DCL-98-154 e, ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.13-5 APPLICABILITY: DC REQUEST:

ITS Bases 3.4.13 LCO and Bases SR 3.4.13.1 (Diablo Canyon)

Comment: The discussions include CRDM canopy welds as exceptions to the definition. That exception is not included in the Bases discussion for ITS 3.4.13 Actions B.1 and B.2 and the exception is not justified.

FLOG RESPONSE (original): LCO 3.4-13 is intended to identify " impending gross failure" (CTS Bases 3.4.6.1) where as leaking seals and gaskets are recognized as not being associated with impending gross failure. The CRDM canopy welds are specialty seals where the " strength is provided by a separate device"(ASME Section Ill,1989, NB-4360). The function of this weld is to provide a seal against leakag, rather then provide reactor coolant pressure boundary integrity against gross failure. Leakage of a CRDM canopy seal weld is not indicative of impending gross failure of the pressure boundary. They should therefore be included as IDENTIFIED or UNIDENTIFIED LEAKAGE and not as PRESSURE BOUNDARY LEAKAGE.

FLOG RESPONSE (supplement): On October 8,1998, the NRC requested supporting documentation that leakage of a CRDM canopy seal weld is not indicative of impending gross failure of the pressure boundary. Attached is documentation to support this position.

NJ ATTACHED PAGES:

Supporting Documentation PG&E Letter to NRC (DCL-89-060) dated March 10,1989 Westinghouse letter to PG&E (PGE-88-622) dated June 14,1988

,O

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,. .y y, , . ,, ..,, -.g?. -u ar. M v">n;1.; W .+i; C + . * D'*~UT d' I ff 2 wv.' '. 'U@(Wkg f.

.s w)$,O.

ii4646p' 77h PowerSystems N .

.i e

aYmre orpsmien g jgl*** 4

. aos 355 hasewy Pomse aru mn:nt 1 Str. J. D. atiffer C* ' >

Vice Presiderst, Raclear Ptasar Generation l pacific Gas & Electric company June 14,1988 ,

77 anale street 35-CPIA-CPIr II-88-394 aan Francisco, GL 94106 Attenticm3 T. L. Grebel IRCIFIC Gh8 Ne 1!m ritIC CG9NfY MX1 EAR PLANT, DIABID OmW S11E 13 TITS 1 AND 2 O' wrw e m sum- w w w -w mvum i j

n. Daar str. aniffers >

o ;

a this 1steme is in zuspense to a reganst by ycnar 30s. J. R. Minds of the Diablo

v. Regulatory cupliance group to fr==alk a Technical 5, Anterpretation anda recently by Westingneuse. --

p walds an the Gest penetrations need not be consideredof boundary the inew is to s

lankage as defined in the Diablo Carwen Technical basis for seking this interpretatism is that the annepy anal wald is a f h tions. The ;

ncrMstre-=1 lapendin vald musti that lanka 4

h. a di==g% grams structural failure.ge frca it would not be indicative ofShe f ccup of the bases is attached to this latter.

M j

g Mota that this interpretation does not apply to OEM annepy seal walds that have had lankage repaired by a wald overlay or multiple wald build-tp peccans.

] 1 If you beve arqr gasstians concerning the information in this letter please contact A. N. sicari at (412) 374-5585 ce the undersigned.

y.

(

Very truly yours, WWF22H3EDER EEECI5 TIC C3tRRA2TGi y

}fh $

J. C. ,lemager '

Pacific Gas and Electric Project A. M. Sizari Att w+ - rit:

Interpretation of the Tech spec definition of stessutt 80GEmitr 1EREAGE as it applies to OtDt canopy seal walds. -'-. , '

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.f, v Y 114646 At:tactment to RE-44-422

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RR7ECrt Tectatical @ecification Interpretation ninhin Cargon Unic Hos.1 & 2 section 1, Definitions Phia 1.24,

nasasE amm IEMUm DJEIOf Is laskege from a tapt penetration canopy meal wald to be construed as NumEEE 30DEMtf IENOW cr as I!ENITFIED e tamINITFIED LENSCC as these are defined in the Tacimical specificaticms for the purposes of W imae= with 100 3.4.6.2, Reactor Coolant Imakage.

INIERSWEUtI213f1 Imakage fries a OM i .a tion canopy seal wald should be construid as

~7 either IIENITFIED cr 13tIZENIZFIED IEN0m and not as W.ma PCKNM 1EMGM for the gaarposes of cumpliance with IID 3.4.6 2, Anactor Coolant tankage. This is consistant with both the language and the intant of the CD s@ ject Tactusical specifications.

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

?' Sha Csut penetration canopy seal wald is not a structural wald and ttnas is not reliad upcm to for maintaining the structural integrity of the reacter coolant system. Sha Cast is attamed by anons of a threaded joint usie J) ' prwides the seennical means of holding the act was in place een

% the system is preneurised. mi=itsely, In the case of spare estat n penetrations, a threaded plug is installed with the threaded joint prwiding the machenical manns of closing the reacter coolant syntaa.

Secticn 1.24 of the Suchnical specifications defines .1I4tEB8t3E RIBEDMtf IEMom as lankage thecups a norwimalahim fault in a Beacter coolant systen w- bo#, pipe wall, ce vammel wall." The Capt penetration canopy seal wald is claerly not covered by the language of this definition.

Shis ir% ^= "an is also consistent with the intent of the subject Technical Wfimtion. Sectica 3/4.4.6.2, BMES for CFEUG2 GEL IEMas, states that the ressen for gurchibiting PRESENE EMNtf 12 Nom of ary negnitude is that such lenkega any be indicative of an ingending grams failure of the preneurs boundary. The canopy seal weld pewidas a asal against any lankage Wich might otherwise cocur 'Aw:. the threaded joints and leakage throuq$1 canopy heal walds is in no way irstkutive of ary 4Wi'ig gross failure of the reacter twelant preneurs bcondsc.y.

Enclosura 2 PG&E Letter DCL-98-154 f~

()) ADDITIONAL INFORMATION COVER SHEET ADDITIONAL iWFOPMATION NO: O 3.5.5-1 APPLICABILITY: DC,CP REQUEST:

Section 3.4 DOC 6-21 LS 35 Section 3.5 JFD 3.5-4 CTS 3.4.5.2 Action b (CP)

CTS 3.4.6.2 Action b (DC)

ITS 3.5.5 Action A This change is a change to both the CTS and the STS and is ' c9 yond the scope of the conversion review and is generic. DOC 6-21 states that this change is consistent with WOG-84.

Comment: Please provide the current status of WOG-84. If WOG-84 is not approved by the TSTF, then this change should be withdrawn from the conversion submittal at the time of the TSTF rejection. If WOG-84 has not been acted on by the TSTF, or is approved by the TSTF but not approved by the NRC by the time the draft safety evaluation is being prepared, then it should be withdrawn from the conversion submittal at that time. This change will not be reviewed on a plant-specific basis.

FLOG RESPONSE (original): DCPP and CPSES will continue to pursue the revisions (O) proposed by this change. WOG-84 is now TSTF-236 which was approved by the TSTF on 1 February 5,1998. The NRC has requested that the WOG provide additionaljustification to i support the extended Completion Time and changes to Required Action A. The WOG is l preparing that information in addition to proposed changes to the 3.5.5 LCO and SRs. The revised traveler will be issued in the near future to the NRC.

l FLOG RESPONSE (revised): Per discussions with the NRC, since TSTF-236 has not been approved, the extension in seal water injection flow AOT from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months associated with the TSTF will be withdrawn.

ATTACHED PAGES:

Attachment 10 - CTS 3/4.4 Enci 2 3/4 4-19 i Enci 3A 10 )

Encl 3B 9 l

Enci 4 Table of Contents,62,63 I l

Attachment 11 -ITS 3.5 ,

p) t EnclSA Enci5R Traveler Status sheet,3.5-10 B 3.5-37 Encl 6A 1 Encl 68 1

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION G

3.4.6.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE. g g ,. g g , g . 5 ,

b. 1 gpm UNIDENTIFIED LEAKAGE. ka * *:P ^N e c

5' e ;

cGy ,;gg-y [ggg[, .jg,3qqggy_}@gg_ty,ggg

. . . , y - _ , , _ . _ _

h Waton and d.10 gpm IDENTIFIED LENAGE from the Reactor Coolant System, e- 40 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 06-06 A

+ 20 psig.. and, I

f. I om, gpm--'^2ka c ,J c at m,,.m.

,.o a heterormer Ccobnt 4":t^-'

- r-, 6 ,mm0,e::ur^

-r,mm, ofsr 2235 rmmm, *20

<m m p:@" 'er m 06-07.to I N5 [E5k5gh frbE eaci Riacioblo6Ia6U Systim Pfessdre is61ation" l 06-25-LS26 Valve shall>be < 0-5 gpm per nominal inch of valve. size u of 5 gpm at an RCS. pressure > 2215 psig;and 5 2255 psig. p to.a maximum APPL ICABILITY: MODES 1.2.3b.nd4#.b N S9 l ACTION:

a. With any PRESSURE B0UNDARY LEAKAGE. be in at least HOT STANDBY withb 6 l hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With any Reactor Coolant System leaka e reater than any one of the above

[ limitsm excluding PRESSURE BOUNDARY L GE. reactor coolanti seal injection flow.cand leakage from Reactor Coolant' System' pump'(RCP) pressure isolation valves, reduce the leakage rate to within limits within 4 06-09-LS10 hours or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

(new)? ht ;06%1 S.

"/T th owRCP_ seaFinfectionsflow~oreater2tharf sou1 valent /co a sinaleAWExm e an curalna tietabovelimitilpfify Traln is/avai >hYe a J-Wittrfn '4 1ours/andfreduce the flaw- rate +toxwithin -:IlmitstwitningN'1 f__._i i--i hours or X! in at .1 east' HOT STANDBYJwithin;the next>6' hours;and in R7T ' ' ' ' ' ' ~ ~ '

^

SHUTDOWN within;the.following 6:h.ours! ' " " ~

06 09-LS10 ,l

c. With any Reactor Coolant System pressure isolation valve leakage greater than the above limit.1solate the high pressure portion of the W- '

affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one Na closed manuaF and/w deactivated automatic or check 06-11-LS11 valve #, and within'72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months b d series: closed manual, deactivated automatic.yttheJuse:of or check! valve- aes or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and'1n COLD ~SH OWN within the 0612.M following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .@(da w j es.29.c3m 6 0 - 3o - A For MODES 3 and 4, if steam generator water samples indicate less than the minimum detectable activity of 5.0 E-7 aicrocuries/mi for principal gama emitters, the leakage requirement of specification 3.4.6.2c may be considered met.

~

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% 29- LM G r Septmk. Ac+ico en+ry is ancused Sv a;ch PW PM P&

Q

_ggg an Env agnan oneoerwe, mv cmans n Regu.m ws y nsus m ;~pmee by w30 l 1

DIABLO CANYON - UNITS 1 & 2 TAB 11.3 3/4 4-19 ?c"r :~ ";: , 'jy -

DESCRIPTION OF CHANGES TO TS SECTION 3/4.4 I

CHANGE hlUMBER NSHC DESCRIPTION 16-21 M (TiiIs change increa s the RCP sealin . on flow Complet n Time from to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , with new added verifi on that at least 10 4 of the assumed cha ing flow remains a ilable. The Bases f the sealinj tion flow limit r tes to ensuring ade ate charging flow d nng post-LO injection he revised ACTIO continue to assure is basis is adeq tely addressed by pr iding an ECCS-like equired Actio ITS Sp ification 3.5.2 allows 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion me for one or ore CS subsystems inop ableif atleast100% the assume CCS flow is available. The sea njection flow ACTION have been m ified so that if the remaining ch ging flow (with some i perabilityin th charging system)is great than or equal to 100% the assumed st-LOCA charging flow, hours is allowed to ree re _OPERABILI This changej

[s consisten th Industry Traveler W

-84 ,) {gg j3eg }

06-22 M Not applicable to DCPP. See Conversion Comparison Table (Enclosure 3B).

06-23 LS25 CTS 3.4.6.1," Leakage Detection Systems",is revised such that the provisions of Specification 3.0.4 are not applicable. This will allow entry into the applicable MODES with only one of the Leakage Detection Systems OPERABLE, subject to the requirements of the ACTION statements. This change is consistent with NUREG-1431 and Industry Traveler TSTF-60 and is acceptable because of the diverse available to detect RCS leakage. (Inser+l (q

'] 06-24 M G 3ff.iS- I Not applicable to DCPP. See Conversion Comparison Table (Enclosure 3B).

06-25 LS26 The Operational Leakage LCO has been modified to change the allowed leakage limit for RCS PlVs for consistency with improved TS SR 3.4.14.1.

The RCS PlV LCO permits system operation in the presence of leakage through valves in amounts that do not compromise safety. (Inscr+)

06-26 LS30 The CTS surveillance requirement for performing a RCS water inventor?

balance is modified to allow deferral of the water inventory balance such that it would be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving steady state conditions. The RCS water inventory balance must be performed with the reactor at steady state conditions as discussed in the ITS Bases. This change is in conformance with industry Traveler TSTF 116, Rev.1.

06-27 A Not applicable to DCPP See Conversion Companson Table (Enclosure 38).

06-28 LG Not 'pplicable to DCPP. See Conversion Companson Table (Enclosure 38).

07-01 R Not applicable to DCPP. See Conversion Companson Table (Enclosure 38).

%-29 LS E Irm+)

s p-t A g.4 @ #N V

DCPP Desenption of Changes to Current TS 10

v (__) 'd CONVERSION COMPARISON TABLE - CURRENT TS 314.4 Page 9 of 15 TECHNICAL SPECIFICATION CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO COMANCHE WOLF CREEK CALLAWAY CANYON PEAK 06-16-LS This change removes the requirement for monitoring the Yes Yes Yes Yes 14 reactor head flange leakoff system.

06-17-LG The definition of steady state is moved to the Bases. Yes Yes No, WCGS did No, Callaway not have this did not have definition. this definition.

06-18-LS This change relaxes the requirement for PlV testing No, MODE 5 Yes Yes No, already in 15 fo!!owing operation in MODE 5. The previous testing CTS per requirement was testing following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in MODE 5 requirement is Amendment which is revised to 7 days in MODE 5. not part of CTS. 35.

06-19-TR This change removes the specific requirement for Yes Yes Yes Yes 3 performing the PlV surveillance prior to returning a valve to service following maintenance, repair or replacement 06-20-A IST requirements are moved to Administrative Controls Yes Yes No, WCGS No Callaway Section 5.5.8 of the improved ITS. does not have does not have this this requirement. requirement.

This ch ge increaseMhe RCP se injection flow -Ves NA Yes AM Me, see CN 5- Mc, ccc CN 06 06-21 (9- t-G VA Comp tion Time fro 4 to 72 ho s, with a ne dde 28-LG ev4 serif' ation that at ast 100% the assume arging

[lo remains av able. J-e Akg %

06-22-M This change adds a new ACTION to isolate the affected No, not part of Yes Yes Yes RHR penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months if the RHR suction current DCPP isolation valve interlock function is inoperable. TS.

06-23-LS The leakage detection system specification is revised Yes No, the non- Yes Yes 25 such that the provisions of 3.0.4 are not applicab!q/- applicability of g O/x Vwo.%i& sv*ms cco be ,ncP**Y- 3.0.4 is already ( q 3. v < 5-/

wief memJ g 3 c.1 part of the CTS. ' -

06-24-M Revises ACTION to require going to COLD SHUTDOWN No, the 600 No, the 600 Yes Yes rather than HOT SHUTDOWN with an RCS pressure psig ACTION is psig ACTION is less than 600 psig. not part of the not part of the CTS. CTS.

DCPP Conversion Comparison Table - Current TS

T L S-3 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 6

-33........................... . . . . . . ............58 L S-34 . . . . . . . . . . ................... . . . . ............60 _ _

LS-35............................... . . . IbfPped.... @ (6 ss@

4 L S-3 6 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 64 LS W 4pikcadele M 93.q,y,3 LS 36 %

V. - Recurring NSHCs i T R -2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . J6 G 8 T R -3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . S769 1

i 1

i l

l s )

1 d

J J

4 l

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IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATIONS s

/] NSHC LS35

( ,/ 10 CFR 50.92 EVALUATION FOR SPECIFIC LESS RESTRICTIVE TECHNICAL CHANGE -

M This change increases th seal injection flow Completion Time from 4 to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , with new added verificaton that at least 100% of the assum charging flow remains available. The Bases for seali ction flow relate the limit to ensunng adequate charging w dunng post-LOCAinjection. The revised ACTIO continue to assure this basis is adequately addressed by provt g an ECCS-like Required Acton. Specificatio .5.2 allows a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time for one or more ECCS subs tems inoperable if at least 100% of the as med ECCS flow is available. The sealinjection flow ACTIONS have en modified so that if the remaining ch ging flow (with some inoperability in the charging system) is greater than equal to 100% of the assumed po -LOCA charging flow,72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is allowed to restore OPERABILITY. This hange is consistent with indu Traveler WOG-84.

This proposed TS change has been evalu ted and it has been de mined that it involves no significant hazards consideration. This determination has bee edormed in accor nce with the enteria set forth in 10 CFR 50.92(c) as quoted below:

"The Commission may make a final determin ion, purc ant to the procedures in 50 91, that a proposed amendment to an operatinglicense for a facilit licen d under 50.21 (b) or 50.22 or for a testing factitty involves no signtlicant hazards consideration, of operatio of e facility in accordance with the proposed amendment would not:

1. Involve a significantincrease in the probab tty consequences of an accident previously evaluated; or

2. Create the possibility of a new or differe t kind o ccident from any accident previously evaluated; or

3. Involve a significant reduction in a m igin of safety."

U The following evaluation is provided for e three categorie of the significant hazards consideration standards:

1. Does the change involve a signi cant increase in the obability or consequences of an accident previously evaluated?

The proposed change revise the completion time for r toring sealinjection flow from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

The basis of this completio time is to ensure availability f the assumed post-LOCA charging flow. To compensate for the increa ed completion time, a new r irement is added to verify, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , that at least 100% of the assum post-LOCA charging flowis a ilable. Since the change continues to ensure 100% of the assumed c rging flow is available, the propo ed change does not involve a significant increase in the probability or con equences of an accident previousi evaluated.

2. Does the change cre e the possibility of a new or different k d of accident from any accident previously evaluated?

There are no hard are changes nor are there any changes in t method by which any safety-related plant system performs i safety functon. Since the change continues o ensure 100% of the assumed charging flow is available, o new accident scenanos, transient precursors, ilure mechanisms, or limitng single failures are intro uced. Therefore, the proposed change does not c ate the possibility of a new or different kind of accident rom any previously evaluated.

A DCPP No Significant Hazards Evaluations 62

IV, SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATIONS O NSHC LS35 gg

() (Continued)

3. Does this change involve a signifi t reduction in a margin of safe The proposed change does not affect the eptance criteri or any analyzed event. There will be no effect on the manner in which safety limtts or limiting eti sys settings are determined nor will there be any effect on those plant systems necessary to assure - ccomphshment of protection functions. Since the change continues to ensure 1007c of the assume ha ing flow is available there will be no impact on any margin of safety.

NO SIGNIFICANT H RDS CONSIDERAT DETERMINATION Based on the above evaluation, itis con uded that the activities associated th NSHC "LS 35" resulting from the conversion to the improved TS form atisfy the no significant hazards considQtion standards of 10 CFR 50.92(c);

and accordingly, a no significant h ards consideration finding is justified. N g

DCPP No Sigc:ficant Hazards Evaluations 63

1 0 Industry Travelers Applicable to Section 3.5

[g,i.e TRAVELER # STATUS DIFFERENCE # COMMENTS TSTF-90, Rev.1 Incorporated 3.5-6 A n w k; Ncc.

TSTF-117 h 2 3

Incorporated 3.5-1 beea W sec.

TSTF-153 Incorporated 3.5-8 4mm 8 sc.3 TGTI-155 - Not M00rpc ;tcd -WA- Nvi NRC appivveu as d ; e.elm. vui-vii dee.

I orporated 3.f [ DCP and E l

l l

l l l l

i 4

(

q

Seal Injection Flow 3.5.5 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.5 Seal Injection Flow LC0 3.5.5 Reactor coolant pump seal injection flow shall be 56 403-gpm a with[ccrt-4fug:7 charging pump d::ch rge headec3 RCS pressure 2-E2480-2215 psig and 52255 psig3 and the Echarging flow 3 355 control valve full open.

APPLICABILITY: MODES 1, 2 and 3.

l ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Seal injection flow not 'h.1Ve fy 2100% - ow 4) fours] (3f i j wtthin limit- uivalent't a. single -

PERABLE;E -chargin train;isi ailable j A.@ Adjust manual seal hours

' injection throttle valves to give a flow within limit Ox with[ centr 4fug:1 Charg M9 3.55 pump discharge header] RCS pressure 2-[2'80/2215psig

and s2255Fpsig andthe B

{ charging flow} control valve full ~open.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. MQ B.2 Be in MODE 4 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months l-

. C '\

GI .

DCPP Mark-up of NUREG-1431. Rev. 1 3.5-10

m . _ _ _ _ _ _ _ _ _ _ . _ _ . . _ _ _ _ _ . . _ . _ _ _ . _ _ . _ . _ _ _ _ __ _ ._

-a Seal Injection Flow B 3.5.5 BASES LC0 With the di =h:r;c RCS~ pressure and control valve pos n (continued) as specified by the LCO. a flow limit is establishe h1ch (Tssures that the seal injection line resistance is %

o. . m_,g" Lconsistent m.,.,,,.m.,..with

...... ... the

, , analysi m.. .. 1 ,,m s assumptions.f.,:,,..' 'm..'.t. _a 1 in the :=1 dent :n:!" .

_ ,.61imit as's~uresithat inhen the Thil..~.

RCSdepress'rizestfol=ow:inga9 u l .0CA4and;thefflow to the pump seals >IncreasesEthe/resulting flow;tofthefs'eals will be less"than th.e. _311mitgass_um.edtin tt

-s -- .

cident analysis..

l The '4-it On =:! inj=tir '! :: :^ _in^d "ith th^ CCo ,

di =her;^ he:dcr prc =urc 'imit: =d :r Oper ';;ide =ndition l Of the charging #10;; contro! ';;'"c. mu:t be m^t to r^^d0r the ECCS OPER.^SLE. !' thc = =ndition: 4-are not met the ECCS #10. " .ct bc = : = umed the : =ident =:!y =:.

APPLICABILITY In MODES 1. 2. and 3. the seal injection flow limit is ** r. G - i

dictated by ECCS flow requirements, which are specified for 4

MODES 1. 2, 3. and 4. The seal injection flow limit is not applicable for MODE 4 and lower however. because high seal injection flow is less critical as a result of the lower initial RCS pressure and decay heat removal requirements in Q t#s MODE 4. Therefore. RCP seal injection flow must be limited in MODES 1, 2. and 3 to ensure adequate ECCS g performance.

ACTIONS A.1[nW2[

With the seal injection flow exceeding its limit, the i amount of charging flow available fotECCSi]Mection?to the RCS may be reduced. Under this Condition."' action ^mdst be taken to restore the sealb n:ection? flow to below its limit. utred: 1on:A p onsuresstna witnin A- rs;t remain 1 iava11 e5ECCSlch#gingiflow' withoutfa uming r addit Pfal re)Msp )Uofsthef sumedys-LOCA cha. ng. flow A 002if1 aabilit y beWer fied:b ass riniht >CCPParei RA 41 P" y d red W nn<A? the tal awsJfhe3e o)erator W hours from the time the flo2 above the 11mitJti = ;; < e"a.in: 1;;;; ef 1s known to

4 @ ;; n d parta n te ' c W t ' M .fto correctly position tne manual valves ono tous oe in compliance with the accident analysis. The Completion Time minimizes the potential exposure of the plant to a LOCA with insufficient e injection flow and provides a reasonable time to restore

' seal injection. flow within limits. This time is (EcemweJacen=r'; t";c tr:1ct= Jwith rc: Ort to the Completion Times W CC C G 3.3.L A n w i icf Other ECCS LCC:. it

.: t== cn Oper=1n; experiene rd i: cuf'icient for i

t& ng =rrecti';c =ti=: by Oper: tion: per zr cl

)

4 Qv okee Gcc3 Lccs.]

(continued)

MARK-UP OF NUREG-1431. REV. 1 BASES B 3.5-37

(]

C/

JUSTIFICATION FOR DIFFERENCES FROM NUREG-1431 NUREG-1431 Section 3.5 This Enclosure contains a bnef discussion / justification for each marked up technical char ge to NUREG-1431, to make them plant-specific or to incorporate generic changes resulting from the Industry /NRC generic change process. The change numbers are referenced directly from the NUREG-1431 mark-ups (Enclosure 5A). For Enclosures 3A,3B,4,6A and 6B text in brackets "[ )"

indicates the information is plant specific and is not common to all the JLS plants. Empty brackets indicate that other JLS plants may have plant specific information in that location.

CHANGE NUMBER JUSTIFICATION 3.5 1 This change replaces reference to the " pressurizer pressure

with a reference to the *RCS pressure"in the APPLICABILITY, Required Action C.2, and SR 3.5.1.5. Required ACTION C.2 requires reducing pressureer pressure to less than 1000 psig. However, pressurizer pressure instrumentation does not have the range to read that pressure. Consequently, RCS pressure instrumentation is used. For the purposes of this LCO, the use of RCS pressure is equivalent.

This is consistent with Industry Traveler 117.

3.5-2 Not applicable to Diablo Canyon Power Plant (DCPP). See Conversion Comparison Table (Enclosure 6B).

3.5-3 This change adds the word " mechanical" with regard to throttle valve position stop, consistent with the CTS. These valves have mechanical stops that q) maintain the valves in position for proper ECCS performance.

3.54 This change iner es the RCP seal inj ion flow Completion Tiple from 4 to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , with new added verificatio that at least 100 percept of the assumed c rging flow remains av ~ able. The Bases for sealinjection flow relate th imit to ensuring adequ e charging flow during st-LOCA injection.

The r ised ACTIONS continu to assure this basis is equately addressed by oviding an ECCS-like quired Action. ITS 3.5 allows a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months mpletion Time for 1 or ore ECCS subsystem operable if at least 100 percent of the assume ECCS flow is available. 6he sealinjection flow ACTIONS have be modified so that if the r aining charging flow (wit some inoperabilit in the charging system) ~ greater than or equal to O percent of the ssumed post-LOCA char ing flow,72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is allow to restore OP ABILITY. This change consistent with industry T v le -

h/ [W Useaj Nh 3.5-5 This change deleted reference to CCP discharge header pressure from the LCO and ACTION A to reflect CTS (3.4.6.2.). A description is added to the Bases which provides the methodology for adjusting the seal injection throttle valves consistent with plant-specific analyses.

/m I%l)

DCPP Desenption of Changes to improved TS 1

,--.~

i ( l Q,l G' (Q r')

CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431, SECTION 3.5 Page 1 of 1 TECHNICAL SPECIFICATION CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE WOLF CREEK CALLAWAY PEAK 3.5-1 Replaced " pressurizer pressure" with "RCS pressure " Yes Yes Yes Yes 3.5-2 The Completion Time of LCO 3.5.1, Condition B, is No, not part of the No, not part of the e ice el Ala M Yes, CTS per OL changed from 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to reflect the CTS. CTS. CTS. A end ent endi g.

$[ Amendm

91. G No.

3.5/-2) 3.5-3 Adds the word" mechanical" with regard to throttle valve Yes Yes Yes Yes position stop consistent with the CTS.

3.5-4 (This chan e increases 2 e RCP :> - i sow Go lenon' -Ves NA VeeMA Nc, LOO 3 5 5 b -No, LCO 3X5 is Time fr 4 to 72 ho s, with a w added rification W 'ppWNe. not app'icab!e.

that least 100 p e_nt of th ssumed arging flowj NA NA re ins availabt 4%% -

3.5-5 Deleted reference to CCP discharge header pressure to Yes Yes No, not part of the No, not part of the reflect CTS. CTS. CTS.

3.5-6 SR 3.5.3.1 Note is moved to LCO per Traveler TSTF-90. Yes, per LAR 96- Yes Yes Yes 03.

3 5-7 Not used. N/A N/A N/A N/A 3.5-8 Moves the Notes from the " APPLICABILITY" to the "LCO." No, not part of Yes Yes Yes Also revises the wording in Note 2 from " declared CTS.

inoperable" to "made incapable of injecting."

3 5-9 The seal injection /retum valves (BGV0198-BGV0202) No, not part of the No, not part of the Yes Yes are included in ITS SR 3.5.2.7 since they are included CTS. CTS. ,

in CTS 4.5.2.g.2. l DCPP Conversion Comparison Table - Improved TS

Enclosure 2 PG&E Letter DCL-98-154 ADDITIONAL INFORMATION COVER SHEET

,Os ADDITIONAL INFORMATION NO: CA 3.5-002 APPLICABILITY: CA, CP, DC, WC REQUEST (original): Revise ITS 3.5.4 Bases to indicate that the RWST LCO, by virtue of its temperature, volume, and boron concentration limits, also satisfies Criterion 2 (initial conditions of accident analyses).

REQUEST (revised): Revise various additional ITS Bases regarding the correct application of Criterion 2 of 10CFR50.36(c)(2)(ii). These changes are consistent with the attachment to a May 9,1988, letter from T.E. Murley (NRC) to R.A. Newton (WOG) entitled "NRC Staff Review of NSSS Vendor Owners Groups' Application of the Commission's interim Policy Statement l Criteria to Standard Technical Specifications."

1. Revise ITS 3.5.1 Bases to indicate that the Accumulators LCO, by virtue of its pressure, volume, and boron concentration limits, also satisfies Criterion 2 (initial conditions of accident analyses).

2. Revise ITS 3.5.4 Bases to indicate that the RWST LCO, by virtue of its temperature, volume, and boron concentration limits, also satisfies Criterion 2 (initial conditions of accident analyses).

j p 3. Revise ITS 3.6.7 Bases to indicate that the Recirculation Fluid pH Control (RFPC) System,

'Q by virtue of its TSP-C depth limit which ensures a minimum equilibrium sump pH of 7.1, also l satisfies Criterion 2 (initial conditions of accident analyses). (Callaway only) i

4. Revise ITS 3.7.6 Bases to indicate that the CST (and FWST for DCPP) LCO, by virtue of its water volume limit, also satisfies Criterion 2 (initial conditions of accident analyses).

ATTACHED PAGES:

l Attachment 11, CTS 3/4.5 / ITS 3.5 l

EnclSB B 3.5-4 and B 3.5-31 Attachment 13, CTS 3/4.7 / IT 3 3.7 Encl 5B B 3.7-35 i

v)

Accumulators B 3.5.1 BASES APPLICABLE SAFETY in water volume is a peak clad temperature penalty. Fw ANALYSES (continued) 1:rg^ break Depending onsthe NRC-approved methodology used to: analyze large breaks.~ an increase 'in' Water ' volume saa mayte result inleither a peak clad temperature penalty or benefit. ' depending on downcomer filling and subsequent spill through the break during the core reflooding portion of the transient. The analysis makes a conservative assumption with respect to ignoring or taking credit for line water volume from the accumulator to the check valve. The safety analysis assumes values of

?[515S] 60.8% (836 cubic feet) gallen: and 5[5879] 72.6%

(864 cubic feet) g:11cnr asLread on:narrowTrange; level instruments 2 not' including instrument uncertainty'. l e 21Nfc- M;trubrt inaccuracy'valUc: cf'[5520] gallen

"^ lL92M ;;? ?:. : :.:

sp^:'. *M The minimum boron concentration setpoint is used in the post LOCA boron concentration calculation. The calculation is performed to assure reactor subcriticality in a post LOCA environment. Of particular interest is the large break LOCA. since no credit is.taken for control rod 1 assembly insertion. A reduction-4n below?the accumulator !

LC0 minimum boron concentration would pro' duce a subsequent reduction in the available containment recirculation sump boron concentration for post LOCA shutdown and an increase I in the e mum sump pH. The maximum boron concentration Q

b is used in determining the cold ?ea to hot leg recirculation injection switchover time and minimum sump pH.

The large and small break LOCA analyses are performed at I the minimum nitrogen cover pressuren(502 p:12)(595.5

>siM, since sensitivity analyses have demonstrated that ligher nitrogen cover pressure results in a computed peak clad temperature benefit. The maximum nitrogen cover pressure limit (692 p:12)i(64715ipsig)? prevent:

==hter relief v:he act0 tfbn ~2nd ultim:tc!y_ l r=rve: numulater integrity. provides::marginitocassure ~

inadvertentireljefEvalve@ctuation doesinatioccur.

Theselana19 sis: assumed?pressuresareispecifiedsintheSRs.

Volisestare shownLon:theicontrolDboardlindicatorsias!%

readingsioniaccumulatoranarrow;rangeilevel? instruments?

Adjustments ;to the f analysistparametersi for? instrument 1 inaccuraciesvor otherJreasons;are applief to determine'the acceptance: criteria;used41n:the: plant, surveillance" ' ' '

procedures, iThesecadjustmentslassure the assumed; analyses parame_tersyare maintained. ' "

The effects on containment mass and energy releases from the accumulators are accounted for in e appropriate analyses (Refs. 2 and c43,5.cei The accumulators satisfy riterion 3 of the "9C oc licy Stat = nt. 10 CFR 50.36(c 2)(ii).

Cri4esen 2. and3.

MARK-UP OF NUREG-1431. REV. 1 BASES B 3.5 4 (continued)

RWST B 3.5.4 BASES APPLICABLE SAFETY ANALYSES Steam Gene ratoF Tube ^ Ruotu re " (SGTR)

(continued)

Volume Thd~RWST3olum4lieeded-Eresponse to'h~SGTR'if not"an explicitsassumption sinceithe required; volume is much'less thanithat required:byia!LOCA. '

Boration

~

r Borats ERWSTfwater willibelinjectedlint'o the RCS forba SGTR event. CTheLinsertion:of the(control: rods and the negative reactivity provided by the: injected RWST solution-provides sufficient:SDM during thefinitial4; recovery operationst :10ne of theiinitialio mrator reccvery;actionsufor--this event;is to equalize the RC$lpressurfland the faulted: steam' f ~'

generator pressure .to' minim ze?or; stopithe primary-toi i secondary tube: rupture Lflow and! terminate:safetyLinjection; i Further RCS boration:Wilhbefinitiatedlbyfthe, operator;by -^-

manual (makeup ~to the RCS:

~ '

gener.x 2. css.3 CA).5 co2. '

l The RWST satisfies w ritcre 3 of the PC P01107 BtY0Eent 101CFRj50.36(c)(2)(10.

h_E M p se<+ 3

(

LCO The RWST ensures that an adequate supply of borated water l is available to cool and depressurize the containment in the event of a Design Basis Accident (DBA). to cool and cover the core in the event of a LOCA, to maintain the I reactor subcritical following a DBA, and to ensure adequate I' level in the containment recirculation sump to support ECCS and Gentm -ment Spray Syst e pump operation in the recirculation mode.

To be considered OPERABLE. the RWST must meet the water volume. boron concentration. and temperature limits established in the SRs.

APPLICABILITY In MODES 1. 2. 3. and 4 RWST OPERABILITY requirements are dictated by ECCS and CS Contai m:nt Spray System OPERABILITY requirements. Since both the ECCS and the CS Cont:i m:nt Spray Syste ust be OPERABLE in MODES 1. 2. s

3. and 4. the RWST must Iso be OPERABLE to support their operation. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7. "RCS Loops -MODE 5. Loops Filled."

and LCO 3.4.8. "RCS Loops-MODE 5. Loops Not Filled."

MODE 6 core cooling requirements are addressed by O

(./

LC0 3.9.5. " Residual Heat Removal (RHR) and Coolant Circulation-High Water Level." and LC0 3.9.6. " Residual MARK-UP OF NUREG-1431. REV. 1 BASES B 3.5-31 (continued)

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

(ant FMST)

CST 4 ,

B 3.7.6 l O BASES l 1

requires ~more AFW supplyTthan can be'provided by. the' seismically qualified portion of the . CST.

Thc '"t4ng cycnt for t50 Otherlevents requiring condensate volume 4 are::

1)~ the large feedwater line break coincident with a loss of

'offsite power. Single failures that also affect this event include the following: I

a. Failure of the diesel generator powering the motor driven AFW pump to the unaffected steam generator I (requiring additional steam to drive the remaining AFW i pump turbine); and

b. Failure of the steam driven AFW pump (requiring a longer time for cooldown using only one motor driven AFW pump).

These are not usually the limiting failures in terms of consequences for these events.

a - 3 4 ,. ~. . , 4. - s 4- c e,r, 4r. ,.w.m. v s s m .wwm 4 m. . w, + %. . . ,,

a , ww

4. +. 4. m. ,gm~.,w..+. ww wo. v., .

4. ,,

t

~

2)E ?a break in either the main feedwater or AFW line near where

^ 'the two join. This break has the potential for dumping condensate until terminated by operator action, since the Emergency Feedwater Actuation System would not detect a difference in pressure between the steam generators for this break location. This loss of condensate inventory is partially compensated for by the retention of steam generator inventory. . ._ m. _ cm.

gg - -, -- ~.m..) g, C

The CST satisnes3 riterian 3 of 10 CFR 50.36 (c) (2) (ii).

<-6. s.n. CEE G ~aoE)

LCO To satisfy accident.lo'sgd analysis assumptions, the CST a'nd'FWST must contain sufficient cooling water to remove decay heat 40s ~

[00 60 ~4nute:] following a reactcr trip from 102% RTP, and then to cool down the RCS to RHR entry conditions, assuming a coincident loss of offsite power and the most adverse single failure. In doing this, it must retain sufficient water to ensure adequate net positive suction head for the AFW punps during cooldown, as well as account fc any losses from the steam driven AFW pump turbine. Or Mfc 0 Me4,ti g

te 3 trds W.

The CST level required is equivalent to a usable volume _of a 110.000 3" n:] 41.3% indicated level (164.678 gallons) .

"tcr 1: The FWST level: required is equivalent to a usable volume n

U .

(Continued)

DCPP Mark-up of NUREG-1431. Rev. 1 Bases B 3.7-35

Enclosure 2 PG&E Letter DCL-98-154 O

V ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: DC 3.5-002 APPLICABILITY: DC REQUEST (original): Revise ITS SR 3.5.5.1 by adding a second note that states: "The provisions of specification SR 3.0.4 are not applicable for entry into MODE 3." This note is equivalent to the current technical specification 4.4.6.2.1 c. note except that it does not apply to MODE 4 entry since ITS 3.5.5 does not apply to MODE 4.

REQUEST (revised): Diablo Canyon will no longer pursue this change. It is interpreted that ITS SR 3.5.5.1 Note 1 is essential equivalent to the previously proposed added Note 2.

Therefore, Note 2 will be deleted. .

1 ATTACHED PAGES: ;

Attachment 11 - CTS 3/4.5 / ITS 3.5 Encl. SA 3.5-11 Encl. 5B B3.5-38 O

v

Seal Injection Flow 3.5.5 SURVEILLANCE REQUIREMENTS O .

SURVEILLANCE FREQUENCY SR 3.5.5.1 - - -- ---------- NOTr -- - -----------

ot required to be performed until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the Reactor Coolant System pressure a stabilizes' at a-E2215 psio and s 2255 psig.-}

h7 -

$Nd[..'. .I $. .

Verify manual seal in'jection throttle valves 31 days are adjusted to give a flow within limit with s Ecentr4fu;:1 :b:rging pump disch:rge 5 :dcr]

RCS pressure 2E #40 2215 psig and

s;2255-} psig and the-E charging flow 3-control .

valve full open.

O 1 .

l \

4

!O DCPP Mark up'of NUREG-1431. Rev. 1 3.5-11

Seal Injection Flow B 3.5.5 BASES ACTIONS B.1 and B.2 (continued)

When the Required Actions cannot be completed within the required Completion Time, a controlled shutdown must be initiated. The Comp 10 tion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for reaching .,

MODE 3 from MODE 1 is a reasonable time for a controlled shutdown, based on operating experience and normal cooldown rates, and does not challenge plant safety systems or operators. Continuing the plant shutdown begun in Required Action B.1. an additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is a reasonable time, based on operating experience and normal cooldown rates, to reach MODE 4. where this LCO is no longer applicable. i SURVEILLANCE SR 3.5.5 1

. REQUIREMENTS Verification every 31 days that the manual seal 1rijection throttle valves are adjusted to give a flow " t " below the limit ensures tAa4 proper manual seal injection l thrattle valve position, and hence, proper seal injection I flow, is maintained. The Frequency of 31 days is based on enoineering Judgment and is consistent with other ECCS l vaIveSurveillanceFrequencies. The Frequency has proven !

to be acceptable through operating experience. j As noted. the Surveillance is not requir0d to be perf0r Od et44 completed within'4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the RCS pressure has stabilized ~within T E 20 ^

specified; pressure limits?1; The RCS ran;0 of nord pressure Op0 rating requirement is th specified since this' configuration will produce the required pressure conditions necessary to assure that the manual valves are set correctly. The exception is limited to4hourstoensurethattheSurveillanceistimely.9 REFERENCES 1. FSAR. Chapter 6 and Chapter 15.

2. 10 CFR 50.46.

Srus.urve,IOnc ; fun er modif, d b.jD ble - a % < <a.: +ac e peni.:o n .sf

?. 0,4 Cr ru t- ep b cebot.

S

, 7 mg.

fee y J O

V MARK-UP OF NUREG-1431. REV. 1 BASES B 3.5-38

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

Enclosure 2 PG&E Letter DCL-98-154 i p ADDITIONAL INFORMATION COVER SHEET l

'V l ADDITIONAL INFORMATION NO: DC 3.5-006 APPLICABILITY: DC REQUEST (original): Revise the SR Bases 3.5.2.3 to clarify what is required to verify that the ECCS piping is full of water.

ATTACHED PAGES - additional changes were made to Enclosure SB (' adequately vented" replaces " full of water")

l 1

Attachment 11 - CTS 3/4.5 / ITS 3.5 Encl. 58 B 3.5-17 l

l U

ECCS -- Operating i B 3.5.2 BASES REQUIREMENTS for ECCS operation. This SR does not apply to valves that (continued) are locked. sealed, or otherwise secured in position, since these were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an actuation signal is allowed to be in a non-accident position provided the valve will automatically reposition within the proper stroke time. This Surveillance does not require any testing or valve manipulation. Rather, it involves verification that those valves capable of being mispositioned are in the correct position. The 31 day Frequency is appropriate because the valves are operated under administrative control, and an improper valve msition would only affect a single train. This Frequency las been shown to be acceptable through operating experience.

SR 3.5.2 3 With the exception of the operating CCP centr 4fuga.1 Charging pump, the ECCS pumps are normally in a standby.

nonoperating mode. As such, flow path piping has the potential to develop voids and pockets of entrained gases. I Maintaining the piping from the ECCS pumps to the RCS full of water ensures that the system will perform properly.

injecting its full capacity into the RCS upon demand. This will also prevent water hamer. pump cavitation. and (j~%

g pumping of non-condensible gas (e.g.. air. nitrog hydrogen) into the reactor vessel following an Si signal or during shutdown cooling. The 31 day Frequency takes into i consideration the gradual nature of gas accumulation in the i ECCS piping and the procedural controls governing system i operation. g3 TheTinte'ntiof'the!SR?isito assureitheJECCSipipingtisMW P c1 r;<4Different means:of;verificationEastalternat s 6._.h as verif q fd to -

Ac ven+ hnes of M l emp;ventingitheraccessible loyed;tol provide;thisfassurancgsystem' high.pointssca E d P W .' " W um,. 4 m, ,, s, ,, _ ,4 m cccc m,,s ,~,.,,,4,,, .,

(for non-runn9 I"#8) ngnyL ugg c n;; ; ;s: ;;,.;i z i;; e,7f,.; ,a and accessible hype,e ;;;;' ;,;; ;,;;;;,- 4;;;,;<nq.7,;,;; ;;,g; ; 4;;;

' ~ '~' "

.s,

' ' ' ' " " ~ ' ' ' '

^I ) se c"' 5 nd" ~"

The CCP de:1;r and attached pip 1 g cc-'iguratic" llev: t4 CCP te vent the accumulated ;2:0: "i: the attached cuction and discharg: piping. Continucus venting of +he Cuction piping to the Volume Control tar' 'VCT) :M m ual venting of the discharge pipi"; "1;h pol-t: ::t1 fic: the pump c :ing venti ; cquir: ment: #c- the CC t.

SR 3.5.2.4 O

D Periodic surveillance testing of ECCS pumps to detect gross degradation caused by impeller structural damage or SURVEILLANCE other hydraulic component problems is required by MARK-UP OF NUREG-1431. REV. 1 BASES B 3.5-17 (continued)

Enclosure 2 l PG&E Letter DCL-98-154 ADDITIONAL INFORMATION COVER SHEET ,

ADDITIONAL INFORMATION NO: NR 5.0-001 APPLICABILITY: DC, CP, CA, WC REQUEST: The NRC requested the following:

For the following plants (and CTS sections), the applications identify the CTS l requirements are being relocated to the FSAR: CW (6.2.3, ISEG; 6.5 , review and audit; 6.10.1, record retention); CP (none); DC (6.10.1, record retention); and WC (6.2.3, I ISEG; 6.5, review and audit; 6.8.2.3, procedure changes; 6.10.1, record retention). We discussed relocations to the QA plan with Ray Smith (QA branch) several weeks ago.

The staff needs to have the licensees identify that these requirements are going to the QA plan and thus controlled by 50.54(a). The DOCS for relocating the above CTS i sections are 1-04-LG and 3-09-LG. These DOCS only state the relocation is to the FSAR. The relocation should be to the QA plan.

FLOG RESPONSE: Enclosures 3A and 38 has been updated to reflect the location of subject relocated items.

ATTACHED PAGES:

Attachment 18 - CTS 6.0 i ITS 5.0 Encl. 3A 6 Encl. 3B 7

(~

C

l l

Q(% DESCRIPTION OF CHANGES TO TS SECTION 6.0 (Continued)

CHANGE NUMBER NSHC DESCRIPTION 03-06 A CTS [6.9.1.6], " Annual Radioactive Effluent Release Report" and CTS [6.14c.) are revised consistent with NUREG-1431, Rev.1, to delete the term " Annual" and modify the submittal date. This change provides a reference to 10CFR 50.36a since 10CFR specifies that the report must be submitted annually and include ;

the results from the previous 12 months of operation. <

03 0/ A CTS [6.9.1.5}. " Annual Radiological Environmental Operating Report' is revised to include specific details concerning the i contents of the report. This change is consistent with NUREG-1431, Rev.1.

03-08 A CTS Specifications (6.9.1.7,6.9.1.8 and 6.9.2] are revised to delete the reference to submittal location for the monthly report.

CORE OPERATING LIMITS REPORT (COLR), and special reports. The requirements related to report submittal are contained in 10CFR. Since conformance to 10CFR is a condition of the license, specific identification of this requirement in the TS would be duplicative and is not necessary. Since the plant p requirements remain the same, the change is considered an administrative change. This change is consistent with NUREG-(' 1431, Rev.1.

03-09 LG The record retention requirements are moved to&l.; IE? 9 :nd

,-+(epiementma-omoedtwed The removal of this detail from the

[o. llceasee. 3 CTS is consistent with NUREG-1431. The requirement for Leon %&ct clowme^M retention of records related to activities affecting qualityis I contained in 10CFR 50, Appendix B, Cnteria XVil and other l sections of 10CFR 50 that are applicable to the plant (i.e.,50.71, etc.). Post-completion review of records does not directly assure operation of the facility in a safe manner, as the activities described retaining thesein the documents requirements have_

in;"!:nt already pr;;;Cr;; erJ beer performed. By@

hcensee controlled document $"any changes in these record retention requirements will be adequately controlled under the provisions of 10CFR(SPJS and the applicable regulations.

% ;eo.54- (a.)] '

03-10 LG The Radiation arotection Program is moved to the FSAR consistent with NUREG-1431. This program requires procedures I 1

to be prepared for personnel radiation protection consistent with 10CFR Part 20. These procedures are for the protection of nuclear plant personnel and have no impact on nuclear safety or the health and safety of the public. Raquirements to have procedures to implement 10CFR Part 20 are contained in 10CFR 20.1101(b). Periodic review of these procedures is required by 10CFR 20.1101(c). The CTS is redundant to requirements in the regulation., and thus is deleted.

03-11 A fThe high rad' ~on area is r ised to be con ent with NU GD 1431 and e new Part nontec ical to add requirements rification and orm anges are with NU

(

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O CONVERSION COMPARISON TABLE - CURRENT TS f,.0 Page 7 of 8 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANON COMANCHE WOLF CREEK CALLAWAY PEAK

/1 03-06 CTS [6.9.1.6]. " Annual Radioactive Effluent Release Yes Yes Yes Yes

. A Report," and CTS [6.14c.] are revised consistent with NUREG-1431, Rev.1, to delete the term " Annual" and modify the submittal date.

03-07 CTS [6.9.1.5]. " Annual RadWogk,isi Environmental Yes Yes Yes Yes A Operating Report," is revised to include specific details conceming the contents of the report.

03-08 CTS Specification [6.9.1.7,6.9.1.8 and 6.9.2] are Yes Yes Yes Yes A revised to delete the reference to submittal location for the monthly report, CORE OPERATING LIMITS REPORT, and special reports p@ liceues ca*wied h+}

03-09 The record retention requirements are moved loN Yes - QA Pion Yes -QA Plan a Yes - 9A Plan 6 Yes - CA Pan u LG The requirement m cnw n or.pe, n op %,_ cvea of e ch:pw n o4 we.

for retention of records related to activibes affecting oW FSM . p5 M , USM. FSM.

quality is contained in 10CFR 50, Appendix B, Critena XVil, and other sections of 10CFR 50 that are appitcable to the plant (i.e., 50.71, etc.). N35_C-ool 03-10 The Radiation Protection Program is moved to the Yes' No, deleted from Yes Yes LG FSAR. This program requires procedures to be CTS per prepared for personnel radiation protechon consistent Amendment with 10CFR Part 20. Periodic review of these 50/36 procedures is required by 10CFR 20.1101(c).

03-11 The High Radiation Area section is revised to be Yes Yes Yes Yes A consistent with the new Part 20 requirements.

Changes are nontechnical to add clarification.

03-12 The PCP section is proposed to be moved outside the Yes, move to No, deleted from Yes, move to Yes, move to FSAR.

LG CTS. The PCP implements the requirements of FSAR. CTS per USAR.

10CFR 20,10CFR 61, and 10CFR 71. Amendment 50/36.

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TVA 910 372 6587 March 10, 1989 I PG&E Letter No. DCL-89-060 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk j Mashington, D.C. 20555 '

Re: Docket No. 50-275, OL-DPR-80 l Diablo Canyon Unit 1 1 Licensee Event Report 1-88-004 VOLUNTARY l Control Rod Drive Mechanism (CROM) Canopy Seal Meld Leaks Due to Transgranular Stress Corrosion Cracking Gentlemen:

PG&E is submitting the enclosed voluntary Licensee Event Report concerning CROM canopy seal weld leakage. This report is being submitted for information purposes only as described in Item 19, of Supplement Number 1, to NUREG-1022.

This event has in no way affected the public's health and safety.

Kindly acknowledge receipt of this material on the enclosed copy of this letter and return it in the enclosed addressed envelope.

1 Sincerely, i V' l t' 1 %_ - l

/lJ. %D. Sh, fer i cc: J. B. Martin H. M., Hendonca P. P. Narbut l B. Norton H. Rood B. H. Vogler CPUC Diablo Distribution INPO Enclosure DCI-88->94-N025 N'C " __

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! This voluntary LER is being submitted for information purposes only as desciibed in j Stem 19, of Supplement Number 1, to NUREG-1022.

I On February 25, 1988, with the Unit in Mode 1 (Power Operation), an unexplained increase

! in containment airborne radiation was ob;erved. On March 12, 1988, following plant j shutdown, examination of the reactor vessel head duct work disclosed a leak in the canopy

{

seal weld of the Control Rod Drive Mechanism (CRDN) head adapter plug at spare location

L-5. Subsequent visual inspections revealed additional canopy seal weld leaks at spare

! locations L-9, L-11, and J-5.

From April 8 through April 21, 1988, the identified head adapters were removed anc' replaced with caps welded in place. All repairs were determined to be satisfactory and constituted a permanent repair for these locations.

The metallurgical examinations performed on the head adapters removed from locations J-5, L-9, and L-11, indicated that the leaks were initiated at the inside diameter of the canopy and were caused by transgranular stress corrosion cracking. STP R-8A, " Reactor Coolant System Operational Pressure Leak Test", was revised to include a CRDM inspection.

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. .ca I. Initial Conditions Unit I was in Modes 1 through 6 during this event.

II. Descrintion of Event

. A. Event:

i On February 25, 1988, with the Unit in Mode 1 (Power Operation), an unexplained increase in the Unit I containment airbornt radiation level was observed. Examination of daily containment air samples, both noble gas and radio-particulate, substantiated this increase. However, there was no significant increase in reactor coolant systen leakage as calculated during the regular daily performance of Surveillance Test Procedure (STP) R-108, " Containment Sump Inventory and Discharge - (12 Hrs) - Data Evaluation," and R-10C, " Reactor Coolant System Mater Inventory Balance (72 Hrs)".

The increase in noble gas activity persisted until the end of the refueling cycle. On March 12, 1988, during the refueling outage, higher than anticipated reactor vessel head (RPV) duct work radioactive contamination levels were observed. During the course of investigating these higher levels of contamination on March 12, a leak was observed in the canopy seal weld of the Control Rod Drive Mechanism (CROM) (AA) head adapter plug at spare location L-5. The leak was characterized by deposits of boric acid and rust colored material extending down the CROM housing.

On March 15 and 16, interviews were conducted with the two individuals who discovered the leak. From these interviews, it was noted that the canopy seal weld of the CRDM head adapter plug at spare location L-5 was still weeping at the time of discovery on March 12. No other leaks were visible. Their observations indicated that a minimal amount of boric acid had leaked onto the reactor vessel head insulation. A follow-up remote visual inspection of the Unit I head area revealed a possible leaking weld at spare location J-5.

On March 16 due to concerns about possible similar leaks on Unit 2, smear samples were taken from the Unit 2 CRDH fan ducts. Analysis of these samples revealed that they were consistent with the smear surveys performed on the CRDM fans during the Unit 2 first refueling outage. In addition to smears, daily grab sample data and noble gas data from June 1987 through March 1988 was examined with no increasing trends. These three evaluations provided assurance that there was no similar leakage in Unit 2. -

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0' naa- .= 0 10 0 13 0 17 From April 2 through April 4 Westinghouse personnel conducted a full visual examination of all 79 canopy seals using a Welch-Allyn videoprobt (TVC). This examination confirmed leakage at location J-5 and identified potential leaks at locations L-9, L-11, L-7, and E-7.

Subsequent review of the videotapes by Westinghouse and PG&E personnel determined that in addition to leaks at spare locations J-5 and L-5, welds at spare locations L-9 and L-11 had minute leaks and should also be repaired.

Canopy seal weld at spare location L-7 was identified as requiring sore study and E-7

s determined to have no through-wall leakage.

From April 8 through April 21. spare adapters at locations J-5, L-5, L-9, and L-11 were removed and caps welded on using full penetration butt welds. The

, canopy seal weld at L-7 was later determined radiographically to have no through-wall indications.

The RCS wa:, returned to operating temperature and pressure at the ent of the refueling optage, at which time STP R-8A, " Reactor Coolant System Optrational Pressure Leak Tests," was performed. No additional canopy seal weld leaks were noted.

8. Inoperable structures, components, or systems that contributed to the event:

None C. Dates for major occurrences:

1. February 25, 1988: Event Date-Increase in Unit I containment radiation inels.

2. March 12, 1988: During removal of fan duct work, substantially higher than anticipated contamination levels were discovered.

3. March 12, 1988: Discovery Date-Boric acid discovered on Unit 1 CRON housing at penetration L-5.

4. March 16, 1988: Smear samples were taken from the Unit 2 CRDH fan ducts. Daily grab sample data and noble gas data was examined.

Results indicate that Unit 2 did not have Isaking canopy seal welds.

Visual inspection of the Unit I vessel head confirmsd penetration L-5 was ,

leaking. An additional le.k was discovered at head adapter plug at spare -

location A S.

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5. April 2-4, 1988: Detailed visual inspection of the Unit 1 vessel head identified head adapter plugs at spare locations L-9, L-7, E-7 and L-11 as potentially leaking.

Subsequent review confirmed leaks at L-9 and L-11, and no leakage at E-7.

6. April 8-21, 1988: Head adapter plugs at spare locations L-9, L-11. J-5, and L-5 were repaired using the " cut off and cap" method.

Radiography determined that L-7 had no through-wall indications.

7. July / August, 1988: Canopy seal welds L-9, L-11, and J-5 were metallurgically examined by Mestinghouse and General Electric for root cause determination.

D. Other systems or secondary functions affected:

None E. Method of discovery:

During an investigation of higher than anticipated radioactive contamination levels in the reactor vessel head duct work, a leak was discovered in the canopy seal weld of the CRDM adapter plug at spare location L-5. A follow-up remote visual examination revealed a possible leaking weld at spare location J-5. A detailed visual examination of the Unit I vessel head was performed.

This inspection confirmed the leak at J-5, and identified leaking canopy seal welds at locations L-9 and L-11 as well.

F. Operator actions:

None required G. Safety system responses:

None III. Cause of Event A. Immediate cause:

i 4

The leaks through the CRDMs were caused by cracks in the canopy seal welds of the CRDM head adapter plugs at spare locations J-5, L-5, L-9 and L-11. - i I

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B. Root cause:

On March 14, 1988, an Event Investigation Team (EIT) was forne' to collect and evaluate the pertinent design, operation, installation ano .Npection data required to establish a root cause and to recommend corrective action. As part of this effort Westinghouse and General Electric were contracted to perform metallurgical evaluations on the leaking canopy seal welds at space locations J-5, L-9, and L-11. These metallurgical examinations of canopy seal welds confimed that the leaks were caused by transgranular stress corrosion cracking. The failures were not associated with weld repairs and were not the result of fatigue. It is postulated that the stress corrosion cracking was a result of concentrations of contaminants (chlorides and sulfates) in the stagnant liquid in the canopy annulus and in the crevices formed by the lack of weld penetration.

Chemical analysis of the water drained from the canopy annulus of J-5 verified the presence of chlorides and sulfates.

A further contributor to the failure of the canopy seal weld could be the higher oxygen content suspected in the canopy annulus of the spares. This is due to the canopy seal welds in the spares being at high points of the system.

IV. Analysis of Event The leakage through the canopy seal welds was insignificant. The leakage could not be quantified by the RCS mass balance performed to meet the requirements of Technical Specification 3.4.6.1. The RCS mass balance is considered to have an accuracy of 0.1 gpm. Since the leakage rate was less than 1 gpm as allowed by Technical Specification 3.4.6.1, the condition was bounded by the FSAR accident analysis.

The effect of canopy seal leaks on the structural integrity of the reactor coolant system was also reviewed. The structural integrity of the CROM housing is maintained by the Acme-threaded fastener. The canopy seal weld does not maintain the structural integrity of the reactor coolant system. Even though ASME Section III considers the canopy seal weld a pressure boundary weld, it is not a pressure boundary as defined in the Technical Specifications.

The Technical Specifications define a pressure boundary to be leakage, except steam generator tube leakage, through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall. The definition is further clarified in the Bases for Technical Specification 3.4.6.1. The Bases state that pressure boundary leakage of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. Leakage ;

from a canopy seal weld on a CRDH is not indicative of impending gross failure since the canopy seal weld does not maintain the structural integrity of the -

RCS. Westinghouse reviewed this conclusion and concurred.

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DIO Ol ri O' 0 17 Another effect on the structural integrity of the reactor coolant system is the corrosion of the reactor vessel head due to boric acid. The leaks discovered during the Unit 1 second refueling outage resulted in minimal deposits of boric acid on the vessel head and there was no evidence of corrosion / wastage of the reactor head. In addition, the Unit i vessel head is coated with an aluminum oxide paint that provides additional protection from the corrosive nature of the boric acid.

Since leakage from the canopy seal welds was within the Technical Specification limits and does not affect the structural integrity of the reactor coolant system, the health and safety of the public were not adversely affected by this event.

V. Corrective Actions A. Imediate Corrective Actions:

The CRON head adapter plugs at spare locations J-5, L-5, L-9, and L-11 were removed and replaced with caps welded in place. This modification constituted a permanent fix for eliminating future leakage at the above locations.

B. Corrective Actions to Prevent Recurrence:

1. Shroud inspection access doors and CRDM fan duct air sampling taps were installed in Units I and 2 to allow inspection of the vessel head and provide additional sampling capability. STP R-8A, " Reactor Coolant System Operational Pressure Leak Test," performed during primary system heatup and pressurization after refueling outages, has been revised to include CRDH inspection using these new inspection access doors.

2. Containment airborne particulate, containment noble gas, and the new air sample taps will be used to indicate the possible presence of canopy seal weld leaks. The Chemistry department has instituted a watchguard measures policy to detect primary coolant leaks into containment by utilizing the particulate and noble gas monitors. The Radiation Protection department has drafted a grab sample procedure to utilize the sample taps, which includes directions to notify management if significant increases in radiation levels are noted.

If these three indicators show evidence of leakage, further confirmatory measures should be taken (i.e., direct or remote visual examination).

3. Canopy seal weld leaks in the CRDMs at other plants have been f occurring since the early 1970s. Data on these failures has been _

( compiled through the Westinghouse Owners Group (HOG).

monitoring this effort and will continue to follow the HOG's findings PG&E has been and recommendations to insure that any corrective measures that may be applicable to DCPP are reviewed for implementation.

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l VI. Additional Information a

A. Failed components:

None B. Previous LERs on similar events: !

None 4

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