Text

Proposed Tech Specs Converting to Improved TS Re Section 3.4, RCS, Consistent w/NUREG-1431
	ML20154A290
Person / Time
Site:	Callaway
Issue date:	09/24/1998
From:	; UNION ELECTRIC CO.
To:	;
Shared Package
ML20154A279	List: ML20154A290; ULNRC-03900, Forwards Response to NRC 980715 & 0721 RAIs Re Util 970515 Application for Amend to License NPF-30 to Convert to Improved TS Sections 3.4 & 5.0;
References
	RTR-NUREG-1431 NUDOCS 9810020259
	Download: ML20154A290 (380)
	v • d • e

- . - - - - . - - - . - . - - . - - - - - - _ - - . _ . - - - . . _ ... - .

i RCS Pressure, Temperature, and Flow DE L12its .

' B 3.4.1 ;

y -

l BASES APPLICABLE parameters must be assessed for their impact on the DNBR SAFETY ANALYSES criteria. The transients analyzed for include loss of coolant (continued) flow events and drupped or stuck rod events. A key assumption for the analysis of these events is that the core power distribution m h h 7 - M " Control Bank Insertion Limits": LCO 3.2.311Lxial Flux D1rr renc (AFD)"; and LCO 3.2.4,

{QGadrantPowerTiltRatio )." $ 5.f: Gew /

The pressurizer pressure t- sN lim WD(it of 2269-Wpsig and the RC average temperature limit of WM*F correspond to analytical i limits of 20056 psig and N : - *F used in the safety analyses, with allowance for measurement uncertainty.

The RCS DW parameters satisfy Criterion 2 of tt "",C ";1 icy Statene e m LCO This LCO specifies limits on the monitored process variables - ;

pressurizer pressure. RCS average temperature, and RCS total flow l rate to ensure the core operates within the limits assumed in the safety analyses. Operating within these limits will result in meeting the DER criterion in the event of a DE limited i transient. '

RCS total flow rate Mcontains a measurement error of 2:4--

W based on performing a precision heat balance and using the result to niibr;t; the RCS flow rate indicators.

Potential fouling of the feedwater venturi, which might not be detected, could bias the result from the precision heat balance

in a nonconservative manner. Oar;fer;. ; gralty ;f 0.it for ur.d,.t xted f;;1ir ;f tt.; feed ;ter .;r. turi rei;;; tt.e i.e iral i f .r.: =;;u.

.: ::::=.;; :: :. : for r,; f;u:lri.

Any fouling that might bias the flow rate measurement gr;;ter tter. 0.it can be detected by monitoring and trending various plant performance parameters. If detected, either the effect of the fouling shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling.

The LCO numerical values for pressure, temperature, and flow rate

~

! er; giar, fer tte xser r.t 1;;; tier, but have r.et been adjusted J for instrument. error.

j 9810020259 980924 5 PDR ADOCK 05000483 (continued)

, P Pm MARK UP OF NUREG 1431 BASES B 3.4 2 5/15/97

RCS Minimus Temperature for Criticality I B 3.4.2 i

BASES APPLICABILITY necessary to allow RCS loop average temperatures to fall below (continued) Tm, which may cause RCS loop average temperatures to fall below the temperature limit of this LCO.

i ONS M ,ff g,ff,}

If the pai aseters that are outside the limit cannot be restored, the pla must be brought to a MODE in which the LCO does not app'y. chieve thfs status, the plant must be brought to M00E-3 mammmu within 30 minutes. Rapid reactor d 2 M en~/

shutdown can be readily and practically achieved within a 30 minute period. The allowed time is reasonable, based on operating experience, to reach MODE 3 in an orderly manner and .

without challenging plant systems.

SlRVEILLANCE SR 3.4.2.1 REQUIREENTS RCS loop average temperature ired to be verified at or above 54t 'F every T,,,-T,,, ivietien, & 7,46en-/

l= 1= T.,, :1:= =t ::P. _;.yny=ir .:: in "fS 1::; T,,, : StF.

O.; % ;e ;;difi;; t M 0%. L".;r. ;r.y RCS le;+ ;;;. ;;; t@;retur; i: ' SP'F ;rd tb T,,, T i";i:ti , l= l= T,,, el;,r; i; eierir.;, 0C0 lee,, ;;;r.;; t;;gr;tur;; c.;uld fell kle tk LCO 7;r,;ir; r. wit kut edditic,r.;l c 7,ir.;. The SR to verifv _

RCS loop average temperatures eveM idnutet

- 1;7.:7 W t: ;r ::t tb '::i;rtd ;idti:n Of th S24'2-/

n con +inuously on i.r corvirfen+

svi+L e4 Lee nu Hne Su-voilla ,cor w4reA are +vpic.//v ooe,41,,,dre REFERENCES 1. j'"H FSAR, *vnSe;e tie d ,15..

on e3 ef er rArf+. .rn

. har o Nt/fon#n eo/ -h ,l e .re ne, h A'CE +en,pe n +ure durr a .w ~ ,e n..i. ne l\

(

r..a-hc-rac.de a,r +cn,.~ p.

n.i,y, ~n, c *cakh fr u nselaJ. - -- .v g X

MARK UP OF NUREG 1431 BASES B 3.4 8 5/15/97

RCS P/T Litits B 3.4.3 .

l BASES !

ACTIONS B.1 and B'.2 (continued) i If the required evaluation for continued operation cannot be rcomplished within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the results are indeterminate or unfavorable, action must proceed to reduce pressure and temperature as specified in Required Action B.1 and Required Action B.2. A favorable evaluation must be completed and documented before returning to operating pressure and temperature conditions.

Pressure and temperature are reduced by bringing the plant to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 with RCS pressure < 500 psig $ ?Yf'^~I within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

' redline)

. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

C.1 and C.2 Actions must be initiated imediately to correct operation outside of the P/T limits at times other than when in MODE 1, 2, 3, or 4. so that the RCPB is returned to a condition that has been verified by stress analysis.

The immediate Completion Time reflects the urgency of initiating action to restore the parameters to within the analyzed range.

Most violations will not be severe, and the activity can be '

accomplished in this time in a controlled manner.

Besides restoring operation within limits, an evaluation is required to determine if RCS operation can continue. The evaluation must verify that the RCPB integrity remains acceptable and must be completed prior to entry into MODE 4. Several methods may be used, including comparison with pre analyzed transients in the stress analyses, or inspection of the components.

ASE Code Section XI, Appendix E (Ref. 7), may be used to support the evaluation. However, its use is restricted to evaluation of the vessel beltline.

(continued)

MARK UP OF NUREG-1431 BASES B 3.4 14 5/15/97

~~

"A $ . $ i. . .

Y !* S. Q.$b.) $ 7. m.

$Y r -

'RCS Loops H0 DES 1 and 2

~

c B 3.4.4 BASES APPLICABE i

SAFETY ANALYSES (continued) F 3.f.6 % /

The plant is designed to operate with all RCS loops in operation to maintain DIER above theM during all normal operations and anticipated transients. By ensuring heat transfer in the nucleate boiling region, adequate heat transfer is provided between the fuel cladding and the reactor coolant.

RCS Loops - H0 DES 1 and 2 satisfy Criterion 2 of tt.; =0 P;1 icy M

LCO -

The purpose of this LCO is to require an adequate forced flow f rate for core heat removal. Flow is represented by the number of RCPs in operation for removal of heat by the SGs. To meet safety analysis acceptance criteria for Dee, 3 pumps are required at

+ rated power. - ;

An OPERABLE RCS loop consists of an OPERABLE RCP ir. egr; tic.

p aidir.ii fer;;d 1"a fer t.;;t tier.; pit and an OPERABE SG in l accordarse wit the Steam rator rv ance Program An Xcr/> 0 AfLE i r+ r.r copa eo ery fewere]am{

it *ble h pnvr/e &ced flow. C A-2. +-80+

APPLICABILITY, - In H00ES 1 and 2, the reactorg 4s M criticalg ;r. the has the potential to produce maximus THERMAL POWER. Thus, to ensure that the assumptions 'of the accident analyses remain valid, all RCS loops are required to be OPERABLE and in operation in these MODES to prevent DtB and core damage.

The decay heat production rate is much lower than the full power heat rate. As such, the forced circulation flow and heat sink

., requirements are reduced for lower, noncritical MODES as indicated by the LCOs for MODES 3, 4, and 5.

Operation in other H0 DES is covered by:

LC0 3.4.5, "RCS Loops H0DE 3":

LCO 3.4.6 *RCS Loops H0DE 4":

LCO 3.4.7 "RCS Loops H0DE 5, Loops Filled *:

, LCO 3.4.B.,"RCS Loops H0DE 5. Loops Not filled":

LCO 3.9.5, " Residual Heat Removal (RHR) and Coolant Circulation-High Water Level" (HODE 6); and (continued)

MARK UP OF NUREG 1431 BASES B 3.4 19 5/15/97

RCS Loops MODE 3 8 3.4.5 g > .v BASES ACTIONS M (continued) required RCS loop to OPERABLE status within the Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . This time allowance is a justified period to be without the redundant, nonoperating loop because a single loop in operation has a heat transfer capability greater than that needed to remove the decay heat produced in the reactor core and because of the low probability of a failure in the remaining loop occurring during this period.

M If restoration is not possible within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , the unit must be brought to MODE 4. In MODE 4, the unit may be placed on the Residual Heat Removal System. The additional Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is compatible with required operations to achieve cooldown and depressurization from the existing plant ~ conditions in an orderly manner and without challenging plant systems.

-- { red & Q 3,f[m l If the required RCS cop not ' n operation, and the RTBs-ere cis;d ud Rod rol tem capable of rod withdrawal, the i Required Actio is eit r to re ore the required RCS loop to operation or o d; crac;i;;

all CRDMs(fby openi the RTB or de energizing the motor generator (MG) set . When t REP r; ir, tra cised psitier.

end Rod Control System E capable of rod withdrawal, it is postulated that a power excursion could occur in the event of an inadvertent control rod withdrawal. This mandates having the heat transfer capacity of two RCS loops in operation. If only one loop is in operation,

..a..a; . ;... tt; =; ast r,; ; Fred. The

%.r . .

Completion Times of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to restore the required RCS loop to operation or m ~< ~ de crac;;;; ;11 CL'"is is adequate to perform these operations in an orderly manner without exposing the unit to risk for an undue time period.

(continued)

MARK UP OF NUREG 1431 BASES B 3.4 25 5/15/97

RCS Loops H00E 3 B 3.4.5 BASES ACTIONS D.1. D.2. and D.3 l

(continued)

If twe 1 RCS loops are inoperable or no RCS loop is in operation, except as during conditions permitted by the Note in 6Qsection.

l all CRDNs 7 m;t .5 i crargized by opening the RTB , or de energizing e MG set f S,4:6en-/

g.

All operations involving a reduction of RCS boron concentration must be suspended, and action to restore one of the RCS loops tog g g _f OPERABLE status and operation must be initiated. Boron dilution requires forced circulation for proper mixing, and g,;.717.; tre =s er i-;rar;izir.; tra "C sets

. removes the possibility of an inadvertent rod withdrawal. The immediate Completion Time reflects the importance of maintaining operation for heat removal. The action to restore must be continued until one loop is restored to OPERABLE status and y operation.

SURVEILLANCE SR 3.4.5.1 REQUIREMENTS This SR requires ver41 cation every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that the required loops are in operation. Verification M includet flow rate, temperature, and gl pump status monitoring, which help ensure that forced flow is providing heat removal. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficiert considering other indications and alarms available to the opere: tor in the control room to monitor RCS loop performance.

SR 3.4.5.2 SR 3.4.5.2 requires verification of SG OPERABILITY. SG OPERABILITY is verified by ensuring that the secondary side narrow range water level is :n We - < o"- -

< for required RCS loops. If the SG secondary side narrow range water level is 6 7.f.f" 2

< W4 3 he tubes may become uncovered and the associated loop may not capable of providing the heat sink for removal of the decay hea ., The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view l of other ndications available in the control room to alert the

! operato to a loss of SG 1evel.

MERT $ S.4--24 (continued) t MARK UP OF NUREG 1431 BASES B 3.4 26 5/15/97

RCS Loops H0DE 4

,. B 3.4.6 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.6 RCS Loops H00E 4 BASES -

BACKGROUND In MODE 4, the primary function of the reactor coolant is the removal of decay heat and the transfer of this heat to either the steam generator (SG) secondary side coolant or the component cooling water via the residual heat removal (RHR) heat exchangers. The secondary function of the reactor coolant is to act as a carrier for soluble neutron poison, bori ag The reactor coolant is circillated through ou RCS loops $3.f,6*n-/

connected in parallel to the reactor vessel, each loop containing an SG, a reactor coolant pump (RCP), and appropriate flow, pressure, level, and temperature instrumentation for control, protection, and indication. The RCPs circulate the coolant through the reactor vessel and SGs at a sufficient rate to ensure i proper heat transfer and to prevent boric acid stratification.

In H00E 4, either RCPs or RFR loops can be used to provide forced circulation. The intent of this LCO is to provide forced flow from at least one RCP or one RFR loop for decay heat removal and transport. The flow provided by one RCP loop or RtR loop is adequate for decay heat removal. The other intent of this LCO is to require that two paths be available to provide redundancy for decay heat removal.

APPLICABLE In H00E 4, RCS circulation is considered in the determination of SirETY ANALYSES the time available for mitigation of the accidental boron dilution event. O.c = ;rd "JC bcp; picais this cirsi;ticr..

1 . . .i ;. : . ,. ..- , ,&. , , . ,. - ,.$

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, J , *. . 'l . n ,- e , ii s NI A .

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,7 ,.. =4..'t8..

.. m..... .

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. . .3 ,,, , , * . , . -

.e . .. . 13 ,. . ..i. ,;

, .m. .. . 3 a w.1. , i . . . . . , . -.., e... ..,9,. -

. -........y,q,.

(continued)

MARK UP OF NUREG 1431 BASES B 3.4 28 5/15/97

RCS Loops - H00E 4 B 3.4.6 BASES (continued)

I ACTIONS A.1 m If one required RGG loop is inoperable l sd tw = leep ;r; l i.;gr;ti . redundancy for heat removal is lost. Action must be .

initiated to restore a second RCS or RtR loop to OPERABE status. l The immediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal.

Ibi

f eu rec,;ir;d = lap i; === sd in ;gr;ti;n ud thr;

r; = = 1_s ===, = iusr= = cr = 1mp == u r;;tered t; CO: t;tus te preik ; .;d; dat ns; fer

= => u m r =1.

E L __ ~ ,_~,

k$1t Ast5 bht tolbbh 5 01thi[h[hNr"s [5f.6en-l Bringing the unit to MODE 5 is a conservative action with regard to decay heat reeoval. With only one RtR loop OPERABE, redundancy for decay heat renoval is lost and, in the event of a loss of the remaining RiR loop, it would be safer to initiate that loss from H00E 5 (s 200*F) rather than MODE 4 (200 to 906 *F). The Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a reasonable time, based on operating experience, to reach MODE 5 from HODE 4 in an orderly manner and without challenging plant systems. ;

c-4-end-s-a : .,m If no loop is OPERABE or in operation, except during conditions permitted by Note 1 in the LCO section, all operations involving a reduction of RCS boron concentration must be suspended and action to restore one RCS or RtR loop to OPERABE status and operation must be initiated. Boron dilution requires forced circulation - .. . - < for proper mixing, er.d tk rer;in t; M criticality . - - -< stret at k r;dusd ir, tr.is t3p; cf egr; tier.. The immediate Completion Times reflect the importance of maintaining operation for decay heat removal. The action to restore must be continued until one loop is restored to OPERABE status and operation .

f (continued)

MARX UP OF NUREG 1431 BASES B 3.4 31 5/15/97 11

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

l 1

RCS Loops H00E 5, Loops Not Filled B 3.4.8 _

p, 4 B 3.4 REACTDR COOLANT SYSTEM (RCS)

B 3.4.8 RCS Loops H0DE 5 Loops Not Filled BASES BACKGROUND In H0DE 5 with the RCS loops not filled, the primary function of the reactor coolant is the removal of decay heat generated in the fuel, and the transfer of this heat to the component cooling water via the residual heat removal (R}R) heat exchangers. The steam generators (SGs) are not available as a heat sink when the loops are not filled. The secondary function of the reactor coolant is to act as a carrier for the soluble neutron poison, boric acid.

In H0DE 5 with loops not filled, only RHR pumps can be used for coolant circulation. The number of pumps in operation can vary to suit the operational needs. The intent of this LCO is to provide forced flow from at least one RHR pump for decay heat removal and transport and to require that two paths be available to provide redundancy for heat removal.

APPLICABLE In H0DE 5. RCS circulation is considered in the determination of SAFETY ANALYSES the time available for mitigation of the accidental boron dilution event. ";; "J:" leep; previt.: thi; circul;ti;r.. The g3f; gen _/

flow provided by one R}R loop is adequate for .

heat emovalp

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(continued)

MARK UP OF NUREG 1431 BASES B 3.4 39 5/15/97

RCS Loops MODE 5 Loops Not Filled B 3.4.8 y u BASES ..

APPLICABLE RCS loops in MODE 5 (loops not filled) Mc; M; . iilifid ir.

I SAFETY ANALYSES tM .t C I;11;y St;t- J.; es 1,,;,rter.t ce..tri.',.ter; ti, rid l (continued) r;ictir,r..

LCO The purpose of this LCD is to require that at least two RR loops be OPERABLE and one of these loops be in operation. An OPERABLE loop is one that has the capability of transferring heat from the reactor coolant at a controlled rate. Heat cannot be removed via the RHR System unless forced flow is used. A minimum of one l running RR pump meets the LCO requirement for one loop in

) operation. An additional RHR loop is required to be OPERABLE to meet single failure considerations.

Note 1 permits all RR pinps to be i ;. c,1;d for s M 15 ;;;ist;; J.;r. witdir.; 7c,, era 1;,;.,. t; another. The circumstances for stopping both RR pumps are ' be limited to situations when the outage time is short nd re outlet temperature is maintained > 10'F be ow saturation temperature./ihe Note prohibits boron dilution o raining g7.f,6en-l operations when Rm forced flow is stopped, r-edb.ns Note 2 allows one RE loop to be inoperable for a period of s 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , provided that the other loop is OPERABLE and in operation. This permits periodic surveillance tests to be performed on the inoperable loop during the only time when these tests are safe and possible.

An OPERABLE Rm loop is comprised of an OPERABLE.RE pump capable of providing forced flow to an OPERABLE Rm heat exchanger. Rm pumps are OPERABLE if they are capable of being pc.arwd and are able to provide flow if required.

APPLICABILITY In MODE 5 with loops not filled, this LCO requires core heat removal and coolant circulation by the RHR System.

Operation in other MODES is covered by:

LCO 3.4.4, "RCS Loops HODES 1 and 2":

LCO 3.4.5, ."RCS Loops MODE 3":

LCO 3.4.6, "RCS Loops - MODE 4":

LCO 3.4.7 "RCS Loops H00E 5 Loops Filled":

l (continued)

MARK UP OF NUREG 1431 BASES B 3.4 40 5/15/97

Pressurizer B 3.4.9 BASES ACTIONS A.15 end A.2m (continued) unit must be brought to MODE 3, with 6

...; r, -ter trip L ;;k s ;Fr., withir. 0 Mur; eid to MODE 4- CA .2+4+-

within hours. This takes the unit out of the applicable he[fne\f H00Eher r er.it te egretier, withir. th; herds of O'ff,6en-/

g g,_/ tu asfd :;.;; teres r bx .th1 09.,1ge-e 7 ryHm%p allowed Completion Mea dot tTng experience, to reach the required plant conditirns from full power conditions in an orderly manner and without challenging plant systems.

IL1 If one required group of M pressurizer heaters is inoperable, restoration is required within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is reasonable considering the anticipation that a demand caused by loss of offsite power would be unlikely in this period. Pressure control may be maintained 1 during this time using

J..t.

~.si.

e

~~

ii, C.1 and C.2 If one group of w e. pressurizer heaters are inoperable and cannot be restored in the allowed Completion Time of Required i Action B.1, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.9.1 REQUIREMENTS This SR requires that during steady state operation, pressurizer level is maintained below the nominal upper limit to provide a (continued)

MARK UP OF NUREG 1431 BASES B 3.4 46 5/15/97

i.

I

j. Pressurizer l B 3.4.9 i j u

BASES i SURVEILLANCE SR 3.4.9.1 (continued)

l. REQUIREENTS -

i minimum space for a steam bubble. The Surveillance is performed by observing the indicated level. "; Tr;;a x3 cf 12 Mur;

, sir;; pad, te n-ifying tM pr ~^.a ee2 sift. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months i

interval has been shown by operating practice to be sufficient to

] regularly assess level for any deviation and verify that i operation is wMMft safety analyses j -~Tssumptionk j

O 1 8 6en-/

Alarms are also available for early detection of i abnormal level indications.

1 4

J SR 3.4.9.2 1

i.

l The SR is satisfied when the power supplies are demonstrated to j be capable of producing the minimum power and the associated j

M pressurizer heaters are verified to be at their design rating. This may-be E done by te; tin; tk p ur appi) eutput i 1 ;nd by Rifa;17.; ;; ;intrial ;L2 er, k;ta :1 .t l i ca.tir.;it3 ;;d re;i;t;me. l u The Frequency of ^2 d.;y; . is i j

considered adequate to detect heater degradati -'

'- ' - -- & T,4 6en -/

2rc T. by epic. tin; ;;gui;;; t;, k usstile. l l

redirse) j i.

i fir--+:4-9-3

! ,n __i ___ it_ t__1___ __..___;

i ________1

.rt.2.

. ._. .,m 2.._. . .ry . 2. ._ ._ .t,.. u.. .. .~.... ... r . .~.. 2.,. r,. .. .

j by Cin- 1: pas ;upplin.

i ";i; 0;rveill.x; i--n;trete; tMt th; hatc; s.r. k sr. ally 3 tra.,,f;rred fic, the aani t; th; ;.~. ;;a3 p;;c apply end l u.; sin d. ?.; Tr;;a x3 ef 10 er.tb i; M xd er, e typisi fuel cy;1; r.d i ca.;i;tc..t with ;i;il;r scifistir.; cf ;; r;,.xy p;= upp;i=.

((redIrne)

REFERENCES 1. FSAR, 0;; tie. CllREER O I*N S'"~/

2. NUREG 0737. November 1980.

MARK UP OF NUREG-1431 BASES B 3.4 47 5/15/97

Pressurizer Safety Valves B 3.4.10 y ,

B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.10 Pressurizer Safety Valves BASES C?

BACKGROUND The pressurizer safety valves provi le, in conjunction with the Reactor Protection System, overpres: ure protection for the RCS.

The pressurizer safety valves are tatelly crale;ed pop typaummuseammununni spring loadedg self actuated vetvet 3 @ 3.YIen-/ !

."..e . with backpressure compensation. The safety l

valves are designed to prevent the system pressure from exceeding the system Safety Limit (SL), w psig, which is 110% of the design pressure.

Because the safety valves are tetelly eral;xd ;r.d self actuating, they are considered independent components. The l relief capacity for each valve, 386:49& lb/hr 3 m is based on postulated overpressure transient conditions resulting from a complete loss of steam flow to the turbine. This event results in the maximum l surge rate into the pressurizer, which specifies the minimum relief capacity for the safety valves . - -

- The discharge flow from the -

pressurizer safety valves is directed to the pressurizer relief tank. This discharge flow is indicated by an increase in I temperature downstream of the pressurizer safety valves or increase in the pressurizer relief tank temperature or lev .

Overpressure protection is required in MODES 1, 2, 3, 4, 53

x. . ., c. .

however, in MODE lwithh7.4.6en-/

one or more RCS cold leg temperatures s 275 ~'F end MODE 55 and MODE 6 with the reactor vessel head on, overpressure protection is provided by operating procedures and by meeting the requirements of LCO 3.4.12, - . . , ~ . . . . . . . . . , , , . . ~ . . .

4,". . . . . . . , , , . , . . . . .-

i&4,O The upper and lower pressure ' limits are based on the i it tolerance requirement (Ref.1) for lifting pressures above 1000 psig. The lift setting is for the ambient conditions associated .with MODES 1, 2, and 3. This requires either that the valves be set hot or that a correlation between hot and cold settings be established.

4 (continued)

MARK UP 0F NUREG 1431 BASES B 3.4 48 5/15/97

PressurizIr Safety Valves B 3.4.10 BASES BACKGROUK) The pressurizer safety valves are part of the primary success (continued) path and mitigate the effects of postulated accidents.

$S.f:4 -/

OPERABILITY of the safety valves ensures that the RCS pressure will be limited to 110% of design pressure.

The consequences of exceeding the American Society of Mechanical Engineers (ASME) pressure limit (Ref.1) could include damage to RCS components. increased leakage.' or a requirement to perform additional stress analyses prior to resumption of reactor operation.

APPLICABLE All accident and safety analyses in the FSAR (Ref. 2) that

. SAFETY ANALYSES require safety valve actuatien assume operation of three pressurizer safety valves to limit increases in RCS pressure.

The overpressure protection analysis (Ref. 3) is also based on operation of -

safety valves. Accidents that could result in overpressurization if not properly tensinated include:

a. Uncontrolled rod withdrawal free 3 full power: .

b. Loss of reactor coolant flow:

c. Loss of external electrical load -

d. Loss of normal feedwater:

e. Loss of en . -

AC power to station auxiliaries; and

f. Locked rotor:

E -

E M ( s,[

Detailed analyses of the above transients are contlined in Reference 2. Safety valve actuation >

(above) is r; wired ir, events c. W b M Wbod to limit the pressure increase. Compliance with this LCO is consistent with the design bases and accident analyses assumptions.

(continued)

MARK UP OF NUREG 1431 BASES B 3.4 49 5/15/97

i 1

1 Pressurizer Safety Valves 1 B 3.4.10 BASES APPLICABLE Pressurizer safety valves satisfy Criterion 3 of tre ""O "oMey SAFETY ANALYSES Statement- -- - . -

(continued)

LCO The -

pressurizer safety valves are set to open at the RCS design pressure (500 pi; u - .), and within the ASME specified tolerance, to avoid exceeding the maximum design pressure SL, to maintain accident analyses assumptions, and to comply with ASME requirements. The upper and lower pressure !

tolerance limits are based on the i it tolerance requirements 1 (Ref.1) for lifting pressures above 1000 psig.

The limit protected by this Specification is the reactor coolant I pressure boundary (RCPB) SL of 110% of design pressure. !

Inoperability of one or more valves could result in exceeding the SL if a transient were to occur. The consequences of exceeding the ASE pressure limit could include damage to one or more RCS components,111 creased leakage, or additional stress analysis being required prior to resumption of reactor operation.

APPLICABILITY In MODES 1, 2, and 3 er.d prtiea; ef "00: ? et, eve tt.; L"" er;;r.;

t g reture, OPERABILITY of - valves is required because the combined capacity is required to keep reactor coolant pressure below 110% of its design value during certain accidents. MODE 3 c.r.: pdien; ef "000 ? r; 3 conservatively included, although the listed accidents may not require the safety valves for protection. -

(redline)

The LCO is not applicable in MODE staa ;11 Z: ;;1d leg $ 7,4:6en-/

tc.;.,,;retur;; er; . 27;"i er ir MODE 5 .

~ ,s . ;,, . because L9BP -

+e presided. Overpressure protection is not required in MODE 6 with g reactor vessel head detensioned.

'The Note allows entry into .yagdftWwith the lift settings CA-7A-Odf outside the LCO limits. This permits testing and examination of '

the safety valves at high pressure and temperature near their normal operating range, but only after the valves have had a preliminary cold setting. The cold setting gives assurance that i

4 the valves are OPERABLE near their design condition. Only one (continued)

MARX UP OF NUREG 1431 BASES B 3.4 50 5/15/97

1 Pressurizer PORVs l B 3.4.11 j i.-

BASES ACTIONS 0, ;r.d 3 te griers cy;1ir; 07 tt.; ""; er bled velve; te ;

(continued) acrify tr.eir 0""A"E ;t;ta;. T;;tir.g i; r.et pris, ;d ir, imr m -

l l

M '

~

ally ed either the PORVs must be restored or the flow path l 1solated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The M block valves shoted -

j be closed, but power must be maintained to the associated block l valves, since removal of power would render the block valve I inoperable.

.. l o s.+.n-1 \

1though a PORV may be designated inoperable, it may be able to be manually opened and closed, and therefore, able to perform its function pV inoperability may be due to seat leakagedr.;tr c.;;tica pidi;;;. ;;t~ ;ic cer,tr:1 pidle , or 6 3.+6*M other causes that do not prevent manual use and do not create a possibility for a small break LOCA.

~

For these reasons, the block valve may be closed but the Action requires power be maintained to the valve. This Condition is only intended to permit operation of the plant for a limited period of time not to exceed the next refueling outage (HODE 6) so that maintenance can be performed on the PORVs to eliminate the problem condition.

Normally, the PORVs should be available for automatic mitigation of overpressure events and should be returned to OPERABLE E

. .. ,,.3.n...,..

status prior to entering startup (HODE 2). ;

Quick access to the PORV for pressure control can be made when power remains on the closed block valve. The Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is based on plant operating experience that has shown that minor problems can be corrected or closure accomplished in this time period.

(continued)

MARK UP OF NUREG 1431 BASES B 3.4 57 5/15/97 i

l J

l Pressurizer PORVs '

B 3.4.11

\ -

1 l

BASES ,

1 ACTIONS D.13 and D.2 m ;

(continued)

L If the Required Action of Condition A, B, or.C is not met.-then l tt.; pi st ;ust b; breus,t t; ; = in 2.ict, tt.; LC0 i;s ret

ppl y. T; at,i;;; tt.is ;teta;. the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and 8 to H00E-4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions f /- p orderly manner and without challenging plant systems. In i '" ') ,_ .

sinteir,ing

_ .<~ _ 0 PORY M Cen-/

OPERABILITY mer4e E required. See LCO 3.4.12.

E.1. E.2. E.3. end E.' ,_ J. 1 _ _ - --

If more than one PORV is inope ble and not. capable of being manually cycled, it is necess y to either restore at least one valve within the Completion T me of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or isolate the flow path by closing and removing the power to the associated block valves. The Completion Ti of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is reasonable, based on the small potential for cha lenges to the system during this time and provides the operator me to correct the situation. If one PORV is restored and one V remains inoperable, then the plant will be in Condition B wit the time clock started at the M !

crigin;l l i;1eretie ;f t.;ving ta . . "s ir.ep; retic. If no PORVs are dSA.6en-/

restored within the Completion Time, tt,a tt.; plat est b; !

bre.4,;. to e H00C ir, sich thc LOO t.ecs r.et ; M y. To achicvc l tt,is st;tus, the plant must be brought to at least H0DE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and F to-MBBE4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . i ~> - A x a- ~ ~ -: - -

- ,c.

.g o . - .s ...,ai,. . . . . . . .-. ....m. . ... .,.:,...,.,3.... u ..

o "y . . y.,& ,u - . . ,, -,:4 m. -- .,.,. .: , ; o .. x ,

.- tian,4 ,qu> e 1. b f I. '

, - - 5 ' 4 :' 's 1 #9..- : . . . . ,, s : 4 ' 4

. . . r , .m . ....n .

.. . . . . 2- ..s,, , , .-

The allowed Completion Times are reasonable, based on operating experience.

(continued)

HARK UP OF NUREG 1431 BASES B 3.4 59 5/15/97

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

1 Pressurizer PORVs l B 3.4.11 a w . .,

i BASES ACTIONS E.1. E.2. E.3. end E.4 M (continued) to reach the required plant conditions from full power conditions j jc in orderly manner and without challenging plant systems. In t , w

'f - MODES dSM6,er-/

r 4

m meinte+niet M PORV OPERABILITY g may-be required. See LCO 3.4.12.

(~~ ~$~~

F.1s F.2 r./ r.2

$ 8d I i

7-8 (relline)

If more than one block valve is inoperable, it s necessary to i

eithe restore the block valves within the Comp etion Time of l 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , or place the associated PORVs in manual control and

i. _ restore at least.._,.._.__,o one block valve within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . ;r ' r;;ter; trapf.44en-/

_ - o_,__ <,_m <_..--

The Completion Tises are

reasonable, based on the small potential for challenges to the

! system during this time and provide the operator time to correct the situation.

! .2 y rsAnr- t 2.+-40 & 5.hH-+ j 4

G.15 and G.2m If the Required Actions of Condition F are not met, then the 4

p';nt ;;;t b; hras.t t c. = in his the LCO i,;; 7.et ;pply.

T; nhi;v; thi ;tet;;. the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and " to M8BE-4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

n. .

i

,1 The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. inh 0DE% ~ + . " $ FM6,m/

~

+: meinte+nfet ILITY may-be 3

.- : 4 -

PORV required. See LCO 3.4.12.

i & Ir g . -m ang g

i j (continued)

MARK UP OF NUREG 1431 BASES B 3.4 60 5/15/97

Pressurizer PORVs B 3.4.11 BASES (continued)

SURVEILLANCE SR 3.4.11.1 REQUIREMENTS Block ralve cycling verifies that the valve (s) can be el ,,. . . , . . . .

OA A-

+- <,,..,m,h._

as s r Frequency of 9?. days is the ASME Code, ion XI OM//4 (Ref. S I). 1. ,. .u s,,v+ v._ , o. 4_ , ..,_ w .. __,_1._ ..,s.

u.....

_ 4. . .. .__.. u, . .. , , u. . .4.- .,. - ,

___..1,.. -.-- - - .

3... m uv,vu.n i vi 6. .<,.._._.c,-

"&O t Of @-t=x. be. --- c, ,'r.g it, ble;k ;;h: 2 x x x:ry t: p --it t'.: "a"V te be used fer = ==1 nr.t 01 ef C---'-'t: Ti= +^ ::ter; th; "0"" er.d egr, t'.; bb;k ?..17; t

a. . u. ._ _..__..,......4. .ma m, , . u. . . . u .., ....4.4..,..

_.m,_ ,

. , , ,, m._

, ._ w . u . _

t, _2

...,-........i._._.____.._,o,.,,.

2...,.,

, ._.u.....

__ mm.

m .. - ... - .A

.1 A -E * > L,, LL_ _ ..___J_-. -

.L.,

__L m- L . _-,,,,S.A-m .r . .vv.. . . m v i a

___-..l.,u, 2

e.i. nenA ki. neb _ . .

____ _ "..av

' - . nayvu- a- a .2 vs , - - a-*b' -. -vu- *vsA wa 've- ^^^"srun

^-

bu

__..._1 ,, ,_ ,_ m m_ . - - . >_ > . ,- _

,,..m..,,.. v. .. nw . .. m b . vi .

- vruvww. ... .. u.... -

fulfiiu :.;. 5i0. ps.4: Gen-/

N tar SR - a.

WNote 3 modifies this SR by stating that to M.th 2 gg 3 not required ttr j/' g be met the Required Acti

. wit tggvalve cJos n accordance with

.... ,3 OS.fg

& M.//4 ent"-ReLloe valvein Q 24ll+

rsc}o i-/nn inco ester SR 3.4.11.2 Fftk en uniso!alle /eak

-f),e tc.r sine e +Ae /*oRV fr alre.}y SR 3.4.11.2 requires a complete cy # Operating a PORV through one complete cycle ensures that the PORV can be manually actuated for mitigation of an SGTR. " .: F--';==y ef h/c-3.fooy

= .--th: t beeed cr. : tupke' ef lir cyc

.~.--+ a ...,.+< _ O oe. air,y eqsn';a; nee,;;har ,,, +ha+ :li er.d +Aere ir.dustry Valves usuall <a re,g e.;,, yA,e.rr

,-y -fle'fiir-vguei51,n,f*;lo' r-n,ed ^+ +h* r*p&d Int} ic

..s. ...w c. . . : . :, . , , . . ,. . ,. . . . , . . . - . . , . .

. . i 1 - ; 1 2, y ....'.. . . , . , . ..".,.i..#i ,l _ , , , , E ig i j e C

'. n . .], f.} . n s -. , , ' . . e<. .' ;

..iei.e. s. . . . . : s .; , . ! .gt., ,

, e. E s e,

! N f l . h + pb . lds ,3hb. h. .sve 4 . ejt ta: .

in- " . q83 d f' ,t k :$- M d. s %..

(continued)

MARK UP OF NUREG 1431 BASES B 3.4 61 5/15/97

m. . vs yw -

,. B 3.4.12 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.12 L;w T; ,,;r;tur; Overpr;;;ur; "retectier.

System .~ >

'L","'

m BASES SACKGROLE The L"" Cy;t;; "- controls RCS pressure at low temperatures so the integrity of the reactor coolant pressure boundary (RCPB) is not compromised by violating the pressure and temperature (P/T) limits of 10 CFR 50, Appendix G (Ref. 1). The reactor vessel is the limiting RCPB cv pciwt for demonstrating cuch piotection.

The PTLR provides the maximim allowable actuation logic setpoints for the power operated relief valves (PORVs) and the maximum RCS pressure for the existing RCS cold leg temperature during cooldown, shutdown, and heatup to meet the Reference 1 requirements during the L40F MODES y at djyrne//7 N#C Ru- C - //* 9 in daf.1: 1 CA 74-ess The reactor vessel material is less tough at low temperatures than at normal operating temperature. As the vessel neutron exposure accumulates, the material toughness decreases and becomes less resistant to pressure stress at low temperatures (Ref. 2). RCS pressure, therefore, is maintained low at low temperatures and is increased only as temperature is increased.

The potential for vessel overpressurization is most acute when

_ the RCS is water solid, occurring only while shutdown: a pressure fluctuation can occur more quickly than an operator can react to relieve the condition. Exceeding the RCS P/T limits by a significant amount could cause brittle cracking of the reactor vessel. LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits,"

requires administrative control of RCS pressure and temperature during heatup and cooldown to prevent exceed}ng the PTLR limits.

j';.=f m :..:

This LCO provides RCS overpren(sure protection by having rmini coolant input capability and having adequate pressure relief capacity. Limiting coolant ,..u input capability requires ;11 but er;

. , c .- .. # , , :. . # - . . , . . . . . . . . . .. . incapable of injection f-#dS into the RCS and isolating the accumulators. The pressure relief capacity requires either two redundant RCS lief valves or a depressurized RCS and an RCS vent of suffi ent size. One RCS i

17JJEM" Anm CA-3'.4-do3

,page 83.Hf (continued)

MARK UP OF NUREG 1431 BASES B 3.4 63 5/15/97

l

, ,. . B 3.4.12 BASES BACKGROUM) PORV Requirements (continued)

The calculated pressure limit is then compared with the indicated RCS pressure from a wide range pressure channel. If the indicated pressure meets or exceeds the calculated value, a PORV is signaled to open.

The PTLR presents the PORV setpoints for E4BP- M The setpoints are normally staggered so only one valve -

opens during a low temperature overpressun transient. Having the setpoints of both valves within the limits in the P11.R ensures that the Reference 1 limits will not be exceeded in any analyzed event.

When a PORV is opened in an increasing pressure transient, the release of coolant will cause the pressure increase to slow and reverse. As the PORY releases coolant, the RCS pressure decreases until a reset pressure is reached and the valve is signaled to closs. The pressure continues to decrease below the reset pressure as the valve closes.

RM Suction Relief Valve Rarmiramants h 2A.6*in--/

During L4BP MODES, the idR System is operated for decay heat !

removal and low pressure letdown control. - Therefore, the RIR !

suction isolation valves are open in the piping from the RCS hot 1 1 to the inlets of the PJR pumos. While these valves are (y.elline) ;..d tt.; %0 ;atter, velv;; r.re 7,gr., the RfR suction re11efS 5'.%n-/ i valves are exposed to the RCS and are able to relieve pressure transients in the RCS.

The RfR suction isolation valves 7.r.d tt.; %C extier, velv;; must .

be open to make the RIR suction relief valves OPERABLE for RCS overpressure mitigation. ?.atal a ur; i.f.;rl ake er; c.et gr;itted t; xx; tt.; %0 sectic , i;;istier, v;1v;; t; cir,;;.

The RfR suction relief valves are spring loaded, bellows type water relief valves with pressure tolerances and accumulation limits established by Section III of the American Society of Mechanical Engineers (ASME) Code (Ref. 3) for Class 2 relief valves.

(continued)

MARK UP OF NUREG 1431 BASES B 3.4 65 5/15/97 1

_ . , _ , y , . .

l

m. . , , , . -

% B 3.4.12 BASES BACKGROUlO RCS Vent Reauiramants (continued)

Once the RCS is depressurized, a vent exposed to the containment atmosphere will maintain the RCS at containment ambient pressure in an RCS overpressure transient, if the relieving requirements of the transient do not exceed the capabilities of the venJt Q 3AGa-I Thus, the vent path must be capable of relieving the flow resulting from the limiting MOP ~ mass or heat input transient, and maintaining pressure below the P/T limits. The required vent capacity may be provided by one or more vent paths.

i;r en %'; .;..; te xt tk 1";w ceg;ity r;quirn ..t. it requir;; cavir; e i,re;;urinc ;;f;ti niv;. Tw? irs ; P-Is%

inte...el;. ;;d dintiin;; it; ble;L ain in tM egn p;iti;n. er

i;;il;r13 ;;t;tli;t.ir.;; ; . int b3 egnia;; = Z'; .;c.; nin. ";

..r. Ftt.;;; ;;;nt b; ;La tM 1;ni cf r;;;ter cc,elst. ;; a r.;t t; dr;in tt.; %0 J a ;F .. j (Y

f APPLICABLE Safety analyses (Ref. 4) demorstrate that the reactor vessel is SAFE 1Y ANALYSES adequately protected against exceeding the Reference 1 P/T 07Ab 1

limits. In H0 DES 1, 2, and 3% ;d ..., _ , ..... - ,,.. .y l tgratur; ;n;;dia; C5"T the pressurizer safety valves will l prevent RCS pressure from exceeding the Reference 1 limits. At m .

M and below, overpressure prevention falls to two OPERABLE RCS relief valves or to a depressurized RCS and a sufficient sized RCS vent. Each of these means has a limited overpressure relief capability.

The actual temperature at which the pressure in the P/T limit curve falls below the pressurizer safety valve setpoint increases as the reactor vessel material toughness decreases due to neutron embrittlement. Each time the PTLR curves are revised, the HOP System . must be re evaluated to ensure its functional requirements can still be met using the RCS relief valve method or the depressurized and vented RCS condition.

The PTLR contains the acceptance limits that define the MOP +x requirements. Any change to the RCS must be evaluated against

. .q (continued)

MARK UP OF NUREG 1431 BASES B 3.4 66 5/15/97

-.-.,,r_

, .. B 3.4.12 l

BASES APPLICABLE the Reference l 4 analyses to determine the impact of the change SAFETY ANALYSES on the MOP acceptance limits.

(continued)

Transients that are capable of overpressurizing the RCS are I categorized as either mass or heat input transients, examples of I which follow:

Mass Inout Tyne Transients

a. Inadvertent safety injection: or

b. Charging / letdown flow mismatch l

Heat Inout Tvoe Transients 4

a. Inadvertent actuation of pressurizer heaters:

b. Loss of RIR cooling; or i c. Reactor coolant pump (RCP) startup with temperature as.yumetry within the RCS or between the RCS and steam generators.

The following are required during the MOP . H0 DES to ensure that mass and heat input transients do not occur, which either of the MOP -

overpressure protection means cannot handle: j,,,,,_,_,,,,_, gy,4

a. Rendering 11 Lt one-M E*"

3and incapable of CA-f%

. injectior":

(Ha Ncf f.c sI.r, avahLle das

b. Deactivating the accumulator discharge isolation $a their closed positions: and

c. Oisell ;.;ir.ii '. yH- start of an RCP if secondary

! temperature is more than B'F above primary temperature in

, any one loop. LCO 3.4.6, j "RCS Loops HDDE 4 " and LCO 3.4.7, "RCS Loops H0DE 5 l Loops Filled." provide this protection.

1 (continued)

MARK UP OF NUREG 1431 BASES B 3.4 67 5/15/97

i g

4 5 VI e#J d bb1H

, . . . B 3.4.12 BASES

. APPLICABLE Heat Inout Tvoe Transients (continued)

SAFETY ANALYSES

< . . , ,,,s., ... ., 1 . . . . . . m .. . . . . _ _ , . . .

TS* *'* ^* .

cA.3,44gg Irwhko,,.c on

-l-Le u.te of 4Ae A/ o[,

W

( sccs wq M:" y/of. A

-8 l . , , . . . .

. .. . . ..,. .. g. .......

l '. . . .og.

.. .;m , , , ,

.. .,. ..-(p

3 ..

.....,i.....

. s.. . .. . . : ,'.'.ti.

~

. . ... i...L~ . . . . . , . , . . . #. .,

The Reference 3 4 analyses demonstrate that eith o RS eif 8'n'I valve or the depressurized RCS and RCS vent can Ipaintain RCS pressure below limits when =1; = ll." m erJ M-

..a .vi..e - Q ere is are actuated

. s, the LCO allows only onegg,y'4, 3

"I m erd ne G L.m.u ma q -. - . PERABLE during the 4

(in aUrfro,, do Be N00 (continued)

MARK UP OF NUREG-1431 BASES B 3.4 68 5/15/97

1

m. .r -

.., B 3.4.12 i

BASES I APPLICABLE Heat Inout Tyne Transients (continued)

SAFETY ANALYSES MGP ' MODES. Since neither one RCS relief valve nor the RCS vent can handle the pressure transient s -

. .x4 fra accumulator injection when RCS temperature is low, the LCO also requires the accumulators isolation when accumulator pressure is greater than or equal to the maximum RCS pressure for the l existing RCS cold leg temperature allowed in the PTLR. l The isolated accumulators must have their discharge valves closed and the valve power supply breakers fixed in their open positions. 0; ;xlyn; t tt; citat ;f es;;;leter dixt.;rge i; ;;r ; rari.nr ."00 t-: -l-: ;tur; reg (175'I end bels; trer.

trat of tra LOO '075"I ;ad b;i s).

Fracture mechanics analyses established the temperature of MGP Applicability at NG $

0; ;w:-:-5-: .;;; ;f ; sr.11 bruk in; cf ceel;at x;;d at (LOC')

in L"" "000 4 cafe ; te 10 0~, 50.40 sd 10 0~, 50 '+p;r.dia K i fR;f;. ; ad 0), rc,;;ir;xc.:.; by t; cia; ; ;ai a ;f c, .; lla! p=p ad c,a ;te.gia; m 0"C"'"L ad 0: utueti m := bled.

.,u, o -

I & 2,+. Gen-I PORY Performance O; r~ #

j rtNinc7 The fracture mechanics anal) es show that the ves is protected when the PORVs are set to open at or below the linl t3 shown in the PTLR. The setpoints are derivedbyanalysespiatmodeg performance of the M L'0:' Sy;ts " li;"ia; .

M transient ofih, assuming no _ and CA-3,4 4 3

-6(a A/cf injecting into the RCS M These analyses consider pressure overshoot and undershoot beyond the PORV opening and closing, resulting from signal processing and valve stroke times. The PORV setpoints et or below the derived limit ensures the Reference 1 P/T 11 mitt will be met.

The PORV setpoints in the PTLR will be updated when the revised P/T limits conflict with the L-TOP M analysis limits. The P/T limits are periodically modified as the reactor vessel material (continued)

MARX UP OF NUREG 1431 BASES B 3.4 69 5/15/97

~

B 3.4.12

' BASES APPLICABLE PORV Performance (continued)

SAFETY ANALYSES toughness decreases due to neutron embrittlement caused by neutron irradiation. Revised limits are determined using neutron fluence projections and the results of examinations of the reactor vessel material irradiation surveillance specimens. The Bases for LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits "

discuss these examinations.

The PORVs are considered active components. Thus, the failure of one PORV is asstmed to represent the worst case, single active failure.

RM Suction Relief Valve Performance h *b The RR suction relief valves do not have variable pressee and temperature lift setpoints like the PORVs. Analyses must show that one RHR suction relief valve with a setpoint at or between psig and psig will pass flow greater than that required for the limiting MOP M transient while maintaining RCS pressure less than the P/T limit curve. An =ir; ;11 reli;f

;w r;;;ir ..t; durir; tt.; liaitir; L"' cv.c.t. ;r = ;atier, reli;f v;1;; w'11 sintein = pre;;;r; te within th. v;1;; r;;;d 1"t i;;,,; int, plu; en n;;;1;ti- ___

10t of tt.e 7;t;d lin A' M ** f (, / U u ' T ')

Alth;;;. n;h .

feilar; cc- t;ri;,1

g.im rc..;f v;1v; =3 it nif 2;t ;irgi; ir.;1a;ica er.d le;;tica .;1 thin th =

7^^

, .~

.T,it t . ,.; ;i;^1; f^il;T ;rit;ri^ ' t.;

..,.,...,....,......,....,-a

$hRC5f((T bbs a e d[rease7to NfUe[bloss of O M'b/

toughness in the mactor vessel materials due to neutron t embrittlement 9 Rm suction relief valves must be analyzed to g 7.f,8,p.]

st4 M accosukBate the design basis transients for c HOP-

$dlia)

The Rm suction relief valves are considered active components. !

Thus, the failure of one valve is assumed to represent the worst '

case single active failure. ;

, 'l 1

(continued)

MARK UP OF NUREG 1431 BASES B 3.4 70 5/15/97

i i

i y - m. n c,,.._A_

n . _

, y, B 3.4.12 i

BASES APPLICABLE RCS Vent Performance 1 i SAFETY ANALYSES l l (continued) With the RCS depressurized, analyses show a vent size of E-97 !

j square inches is capable of mitigating the z ~. :ll;;;d LT l

! E acrpressure transient. The capacity of a vent this size is l

! greater than the flow of the limiting transient for the MGP > - l 3 configuration. - ll"I pep c.r.d m l l . OPERABLE. maintaining RCS pressure less than the maximum pressure )

i on the P/T limit curve. !

f j The RCS vent size will be re evaluated for compliance each time the P/T limit curves are revised based on the results of the vessel material surveillance. ;

l The RCS vent is passive and is not subject to active failure.

l The L T Sy;ter - : satisfies Criterion 2 of tr.e ""O "clic/

m , ,. s . . . .

1

. . 1

., LCO This LCO requires that the L T Sy;t a <

is OPERABLE. The L T Systa 4 is OPERABLE when the n+nistse -

calant l input and pressure relief capabilities are OPERABLE.

. 1 j

Violation of this LCO could lead to the loss of low temperature ;

overpressure mitigation and violation of the Reference 1 limits )

as a result of an operational transient.

To limit the coolant input capability, the LCO requires --

. ;r.; ll": pump .

capable of injecting into the RCS and all accumulato ischarge isolation valves 3 closed and 'g"'I

" Tamob111 zed 3 - accumulator pressure is greater than equal to the maxi RCS pressure for the existing RCS cold leg temperature allowed in the PTLR

,. 0W[Q Y$$-2 ,

m.

i.;. : b . ....: :.; . ,; .

, , , , ,; ; . . :7 w, -

. . ,,L:. a... ..

. .. .o .

' ( ., s ' -

v

. 'X. ; t y .) ,' e T s o., o . -

.\.'. .

= * .

~~

.,j g p , . . .a 9g., . . v . . . .. i .;.3. m 3

..--A e... .. -

(continued)

HARK UP 0F NUREG 1431 BASES B 3.4 71 5/15/97

a. wi ,ywwwm

_, ,, B 3.4.12 BASES LCO 3 E- c. > n o -:- m .. .- o4 -cm e , .

(continued) 7: . . , . .. y ,2 . . -..a

.. . c .. - .,

/ (.rM) ggp .

a ..

.. .: a c ., , r . , . - . . . . . ,. - , . - ,,_ ,,,,,,,, ,_ ,,,,, ....,

. , r, . ;x.n,...-

.. .. . . .. 4 -

~ ';. . .. L ' ' - / * * 't g;9. :* '

'. '?~. .

1'

-4 . . -

w: s.3 gs .. . ;.:3 w ,;.;e n..; q32.... :... a..- ...

M E b- A depressurized RCS and an RCS vent.

An RCS vent is OPERABLE when open with an area of :t EM square inches.

Each of these methods of overpressure prevention is capable of mitigating the limiting HOP ". transient.

red lfn e b _

APPLICABILITY This LCO is applicable in H00E E 4_ any RCS cold leg temperature is s EMSF @ (n MODE'4)in H00E 5, and in N00E 6 when the reactor vessel head is on. The pressurizer

$1Mr/

safety valves provide overpressure protection that meets the Reference 1 P/T limits .

Ma O'";"I. When the reactor vessel head is off, overpressurization cannot occur.

.?A.IN m

h_

m E

/ N h_ .._.netsutenememani

- Y

.;7:-

9 (continued)

MARK UP OF NUREG 1431 BASES B 3.4 73 5/15/97

4 M

- b. ,...

p.... B 3.4.12 BASES APPLICABILITY LCO 3.4.3 provides the operational P/T limits for all MODES.

(continued) LCO 3.4.10. " Pressurizer Safety Valves," requires the OPERABILITY of the pressurizer safety valves that provide overpressure protection during H0 DES 1, 2, and 3. ;nd it00C 4 ebev; 275"r. $'?.f,6u-/

t~

Low temperature overpressure prevention is most critical during shutdown when the RCS is water solid, and a mass or heat input transient can cause a very rapid increase in RCS pressure when little or no time allows operator action to mitigate the event.

";; Appli;;bility is ;4ifi;d by e ;tte st; ting ttat enu;;1 uter is;1; tier, is ;nly r; quired u'en the aca;ulater prasur; is r;re tren er et tre =xi u ",00 pr;;sur; ,u_ fer the

.__,,._;t.. i<_ o ,,. , _m _ . . _ . . _ _ o_.._existing _m_ tgr;ture.

mu_

, ud u a a vvywu a#J b4ub I J 5 5 5 35I 4 b bus Tbd. 5 3I 5 e tw bb pl Eug 5 be b3Ub eceu uleter dia;terge iseleti;n v;lV; Surv;illera- t; b; perferad er.ly under ttese pressur; and t ,,;retur; sr.ditki.ts--

ACTIONS -

- - i - .

o. -

n...- ,, , . . . . .

.>, r. . .

, .4 . , . .- .

r. 3, .

V. ) . -l- -l 5% . _

-'}-. 10 ,*L .

A.1 nd B.1 O f Y,6M-/

With twe )B or more HPt . o, . - pumps '

Mim capable of injecting into the RCS, RCS over[ . trization is possible.

To immediately initiate action to restore restricted coolant input capability to the RCS reflects the urgency of removing the RCS from this condition.

." ; quired Action ".1 is sdified by ; 20;; trat peruits tu; irerging pu;ps spebi; cf "CS injutier, fer .15 ;irutes t; ell;u fer pu p su p .

(continued)

MARK UP OF NUREG 1431 BASES B 3.4 74 5/15/97

.t d

4 M

L"" Sy; tee 3

~ ~ B 3.4.12

}

i red lIneb

BASES (continued)

v i SURVEILLANCE SR 3.4.12.1.(SR 3.4.12.2)andSR 3.4.12.3 8 8A,6e/

j REQUIREMENTS To minimize the potential for a low temperature overpressure i event by limiting the mass input capability, a maximum of ex ll"I

~

)

4 pump - .

m are verified M fncapable of injecting into the ;

RCS and the accumulator discharge isolation valves are verified I closed

{ .

24 le;%d c,at.

1 i

[.rM) l

M chargin9 CA-2A-eas re rendered incapable gg g of 1 jecting into the RCS through removing the power from the j pumps by racking the breakers out under administrative control. ,

l a

g A*

i 1 ' '

l

! M A ;1terret; athed of L"" ;,,ntr;l my k gley;d u;in; et 1;;;t tn inde,,,.nd,.nt us; te pre.;nt ; p;;p ; tert ;;h that ; ;ingic f;ilur; er ;in;1; nti; sill ..et 7;; ult in en

! injecti;n int; tk "00. ";i; my M eu,li;Md threa;h tM

! p=p centr;l ; witch kin; pie;;d in pull te leck end ;t 1G;t 7,x ;;.h; in tM dinMrg; 11; pr.th Min; ;iend.

The Frequency of 12' hours is sufficient, considering

. . . m . , . . g . r .. < . , m . other indications and alarms available to the operator in the control room, to verify the

., required status of the equipment.

i 1

m_wA 2

i j Each required RHR suction relief valve shall be demonstrated

OPERABLE by verifying its """ -^"-- - ^ - ' RHR suction

s J

J (continued)

I MARX UP OF NUREG 1431 BASES B 3.4 77 5/15/97 4

i___.____-__-- - _

l l

RCS Operational LEAKAGE l

B 3.4.13

+.

l BASES I i

BACKGR0l#0 consequences of violating this LCO include the possibility of a (continued) loss of coolant accident (LOCA).

APPLICABLE Except for primary to secondary LEAKAGE, the safety analyses do SAFETY ANALYSES not address operational LEAKAGE. However, other operational LEAKAGE is related to the safety analyses for LOCA: the amount of leakage can affect the probability of such an event. The safety analysigs for en event 3 resulting in steam discharge to the atmosphere assumes a 1 gpa primary to secondary LEAKAGE as the 3 initial condition. 4,f/g),

Primary to secondary LEAKAGE is a factor in the dose releases outside containment resulting ( a steam line break (SLB) accident. T; ; 1;;;;r st;at. her accidents or tra ents 8 f f 4

M involveg secondary steam release to the atmospher .uct. ;; e steam generator tube rupture (SGTR). leakage contaminates the sepondary redline ) fluid.

The FSAR (Ref.f) 3 a(nalysis for SGTR assumes the contamina secondary flu' d is caly bri;'ly released via safety .,,-

valvesy';c.d tt.; ;;jerity i; ;tc sd t; tt.; ar.dca;;r. OcQS,f.6en-/

1 ,, i;7i s ry t: x ;.r.d;.ry 'M i; r;lotively ixeangatiel.

We SL:: i; er; li;iting fer ;it; redieti;n rel;;;;;. The safety analysis for the SLB accident asstmes 1 gpu primary to secondary LEAKAGE in one generator as an initial condition. The dose l consequences resulting from the SLB . m accidentg are well within the limits defined in 10 CFR 100 -

er tra :t:ff

.g re;;d lian;ia; be;i; (i.e., a small fraction of these limits).

a. . . .

, .,,,,,y,.. .s The RCS operational LEAKAGE satisfies Criterion 2 of tra OC

." ;1 icy Ot;t;-..;. +e,-s. ..*A (continued)

MARK UP OF NUREG 1431 BASES B 3.4 83 5/15/97

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

RCS Operational LEAKAGE B 3.4.13 BASES (continued)

LCO RCS operational LEAKAGE shall be limited to:

a. Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative 4

of material deterioration. LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB.3

& 2.+.Gm-l LEAKAGE past seals l end gaskets, -- .-o. - ..

is not pressure boundary LEAKAGE.

b. Unidentified LEAKAGE One gallon per minute (gpe) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this LCO could result in continued degradstion of the RCPB, if the LEAKAGE is from the pressure boundary.

c. Identified LEAKAGE Up to 10 gpa of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of Eidentified LEAKAGE and is well within the capability of the RCS Hakeup System.

Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not ce.sidered LEAKAGE).

4 $.M./3-3

~ ' '

'E

~

g . . ..

rp_

7-;_5

~

____ _ _ __m L.""' Violation of this LCO could result in continued degradation of a component or system.

(continued)

MARK UP OF NUREG 1431 BASES B 3.4 84 5/15/97

RCS Operational LEAKAGE

~

B 3.4.13 a . . . ,

. RASES

.r}n)le A CCfeessure, e,,c/we, SURVEILLANCE SR 3.4.13.1 (continued) f(fower / refru,- x s,-

The RCS water inventory balance n t uw - . . m .a m hror m/I'$$*ec er 6 - ,4 %m )

steady state operating condition Pad m;r egr;tir; pr;;;ure. .p/ g ,

Therefore, .

m - o ,.

- this SR is not required to be performed ' """ - ' ' until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of M f)2f./E2 steady state operati m ;r eg retirs pr;;;;r; M v; L a ;;tr.bli;Md.

M I

,- + L(redline) & 2. + Gen -l Steady state operation is regired - - . to perform a proper inventory balanceg calculations during . 4 3.+.6 e /

xx;;;.-irs

. er; at a nful ad ; Lt; r;girs t M irveill;x; te k a t Jar, :tady :t;t; i; st;bli Md. For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the automatic systems that monitor the

, containment atmosphere radioactivity and the containment sump level . It should be noted that' LEAKAGE past seals and gaskets is-not pressure boundary LEAKAGE. These leakage detection systems are specified in LCO 3.4.15. "RCS Leakage Detection Instrumentation."

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Frequency is a reasonable interval to trend LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. /, Lt; ader tk Trcq,.;x.7 als ,

t;t;; tMt thi; 2 i; r;;; ired t; k grierad durir6: t;edy

tet; eg r; tis.

SR 3.4.13.2 This SR provides the means necessary to determine SG OPERABILITY in an operational MODE. The requirement to demonstrate SG tube integrity in accordance with the Steam Generator Tube 1

~

Surve111anc~e Program emphasizes the importance of SG tube 4 J< integrity, even though this Surveillance cannot be performed at l (continued)

. MARK UP OF NUREG 1431 BASES B 3.4 87 5/15/97

RCS Operational LEAKAGE j B 3.4.13 1

1

BASES

SURVEILLANCE SR 3.4.13.2 (continued)

! REQUIREENTS l normal operating conditions, m ,

.?.+. Gen-l 1 1 -

re ore, a ure t urve ance is j considered failure to meet the integrity goals of the LCO and

. LCO 3.0.3 applies. .

,. Y (redline)

REFERENCES 1. 10 CFR 50, Appendix A, GDC - 30. l l 2. Regulatory Guide 1.45, May 1973.

3. FSAR. Section 15

, ~ .

E .

l

g .

1 i

i 1,

I l

l I

I i

4 MARK UP OF NUREG 1431 BASES B 3.4 88 5/15/97

RCS PIV Leakage B 3.4.14 n.c BASES ACTIONS A.1 and A.2 (continued)

The flow path must be isolated by two valves. Required l Actions A.1 and A.2 are modified by a Note that the valves used for isolation must meet the same leakage requirements as the PIVs andmustbewithintheRCPBgerthehighpres;ur;pertion;fthe u'** L.(rejl1ne) Q 2AGen-I Required Action A.1 requires that the isolation with one valve ;

must be performed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Four hours provides time to reduce leakage in excess of the allowable limit and to isolate the affected system if leakage cannot be reduced. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time allows the actions and restricts the operation with leaking isolation valves.

l l

Required Action A.2 specifies that the double isolation barrier !

of two valves be restored by cle;ing ;;;.; etter nin q=lified !

fer i;;l;tien er restoring ea 1shing M PIV b 2- ^~/

(reOInb .- . The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time after exceeding the limit M considers the time required to complete the Action and the low probability of a second valve failing during this time period, er .

7.6 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Oveyletion Ii;; ;Ner G;;; ding tra li;it elleW: fer !

the r;;ter; tion ef the iniing Pr! t; 0~"C ;t;ts. Oi; tie;fr ;. cen;ider; th; ti;; r;@ ired te glet; thi; A;tix ;r.d l th; ici.; preh;.bility of ; ;;sr.d niv; feiling during thi; p;ried.

(R;viser 0;;. Ta eptiens er; praided fer n; wired #ctica A.2. Oc .::;end eptien (72 h;ur ruteretien; i; ;pprepriet; if i;;l;tien of ; ;;;end nin W;uld pix; the unit in en ; arelynd cenditim.)

B.1 and B.2 r'Jk"*} S 3'+'S'"I If leakage cannot be reduced,Lthe sys_ tem isolat_ed, r the other Required Actions accomplished, the plant must be itught to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and H00E 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This Action may reduce the leakage and (continued)

MARK UP OF NUREG 1431 BASES B 3.4 92 5/15/97

RCS PIV Leakage B 3.4.14

o. .

BASES SURVEILLANCE. SR 3.4.14.1 (continued)

REQUIREENTS I i

after Rm is secured and stable unit conditions and the necessary i differential pressures are established.

( GeB Q '*

(SR 3.4.14.f rd T 2.4.14.2 "erify'a; ttat tra "JC stxinur; interlock; er; Of i;f00 craurs trat "00 pronur; sill ret pr;nurin tra "Jn ;y;te; b;yend 100t of it; design pr;nur; ef 000 p;ig. The -

interlock setpoint that prevents the valves from being opened is set so the actual RCS pressure must be < psig to open the valves. This setpoint ensures the Rm design

l. pressure will not be exceeded and the RR relief valves will not l lift.

1 The E month Frequency is based on the need to perform the l Surveillance under conditions that apply during a plant outage, l The 3 month Frequency is also acceptable based on consideration !

of the design reliability (and confirming operating experience) l of the equipment. j 1

l ";;x 0; ;r; r. edified by ";;;; ell;;ia; tra ."JC eut;;1aur; !

fantica t; b; dinbi;d den a;ia; tre "JC Sy;te; ;;;tica reli;f

^4:1J;; fer s .' ;'J;rpren;r; pret;; tic in nGrd;x; sith

= 0.4.12.7. ,

REFERENCES 1. 10 CFR 50.2.

l

2. 10 CFR 50.55a(c).

3. 10 CFR 50, Appendix A, Section V. GDC 55.

i' l 4. WASH 1400 (NUREG 75/014), Appendix V. October 1975.

J l 5. NUREG 0677, May 1980.

1 l 41- -

l h/ C-7.4-0 /0

[ -7v ASME', Boiler and Pressure Vessel Code,Section XI. I 7, -th- 10 CFR 50.55a(g).

I HARK UP OF NUREG 1431 BASES B 3.4 95 5/15/97

,..m___. _ . _ _.___ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ ___

i i

4 RCS Leakage Detection I'nstrumentation m B 3.4.15

( . g..,

BASES .

{ APPLICABLE j SAFETY ANALYSES -

- t -

.r .

5 2

^ '

. . , , ?

l The need to evaluate the severity of an alarm or an indication is important to the operators, and the ability to compare and verify with indications from other systems is necessary. The system response times and sensitivities are described in the FSAR (Ref. 3). Multiple instrument locations are utilized, if needed, to ensure that the transport delay time of the leakage from its source to an instrument location yields an acceptable overall response time.

The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring RCS LEAKAGE into the containment area is necessary.

Quickly separating the identified LEAKAGE from the unidentified LEAKAGE provides quantitative information to the operators, allowing them to take corrective action should a leakege occur detrimental to the safety of the unit and the public.

RCS leakage detection instrumentation satisfies Criterion 1 of the TC Policy Ot;t; at. '- - , -

LC0 One method of protecting against large RCS leakage derives from the ability of instruments to rapidly detect extremely small leaks. This LCO requires instruments of diverse monitoring principles to be OPERABLE to provide a high degree of confidence that extremely small leaks are detected in time to allow actions to place the plant in a safe condition, when RCS LEAKAGE indicates possible RCPB degradation.

";; LCC l: actlsfled ; ;n .e.;ultara ? diverac = auc z a = as es; i;;'. '.;t'; . "; . ;, the iei.tei; ;.1 m;, . en'.te; ir.

ce Lir.; tie With ; ;;;;;;;; er p;-ticul;t; redie;;tivity eai'er qad ; ;entei. cat eir ; eel;r ;;r.dca;;t; ;w ret; ;aiter f S3+68^'I

@'pGJ.dyKjtpp@E previ4;a ;;;;pt;bic ;;;iai=.

(continued)

MARK UP OF NUREG 1431 BASES B 3.4 98 5/15/97

RCS Leakage Detection Instrumentation B 3.4.15

'f i.

BASES ACTIONS 9-1-end-9-e u m * ~....s.- m .

.us (continued)

With the required containment atmosphere: - ~ radioactivity monitor and the required containment eic cooler condensate Mew 1 i r;te er. iter N inoperable, the only means of l detecting leakage 4e - the containment sump M menfie ;.-

m

., i .

This Condition does not l provide the required diverse means of leakage detection.

. .. . . . . . . . .+. ..

l 4

,1

~.1 .

4 The M Required

Action is to restore either of the inoperable required monitert 4

m to OPERABLE status within 30 days to regain the intended leakage detection diversity. The 30 day Completion Time ensures that the plant will not be operated in a reduced configuration for a lengthy time period.

c.1 d : w a .s. .: 3

& 3,A Gen-l i

i If a Required Action of Condition A. B j er-B cannot be met, the plant must be brought to a MODE in whict'the requirement does not apply. To achieve this status, the plart must be brought to at least H00E 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE ! within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The allowed Completion Times are reasonable based on operating 4

experience, to reach the required plant co itions from full

] power conditions in an orderly manner and ithout challenging plant systems. y 4

(continued)

MARK UP OF NUREG 1431 BASES B 3.4 102 5/15/97

RCS Specific Activity B 3.4.16 BASES ,

,~

APPLICABLE The analysis for the SGTR accident establishes the acceptance SAFE 1Y ANALYSES limits for RCS specific activny. Reference to this analysis is (continued) used to assess changes to the unit that could affect RCS specific activity, as they relate to the acceptance limits.

j The analysis is M for two cases of reactor coolant specific activity. One case assumes specific activity at 1.0 yC1/gm DOSE EQUIVALENT I 131 with a concurrent large iodine spike that increases the L101 ;;tivit3 in

- the reactor coolant by a factor of about 50 immediately 1

after the accident. The second case assumes the initial reactor coolant iodine activity at 60.0 yC1/gm DOSE EQUIVALENT I 131 due to a pre accident iodine spike caused by an RCS transient. In

both cases, the noble gas activity in the reactor coolant assumes 12 failed fuel, which closely equals the LCO limit of 100/E yC1/gm for gross specific activity.

The analysis also assumes a loss of offsite power at the same time as the .

SGTR event. The SGTR causes a reduction in reactor coolant inventory. The reduction initiates a reactor trip from a low pressurizer pressure signa er ;n ."CS u.;rtwr;ture oT ;iiin.i.

j The ;;ir.;id;;t loss of offsit causes the steam dump valves to close to protect t ser. The rise in pressure in the ruptured SG disc s radioactively contaminated steam to the atmosphere th the SG p;;r egr;ted M reHet valves ;r.d th; ;;in ;t;;; ;;f;t3 velv;;. The unaffected C 7.t.6co-/

SGs remove core decay heat by venting steam to the atmosphere until the cooldown ends.

l The safety analysis shows the radiological consequences of an SGTR accident are within a small fraction of the Reference 1 dose guideline limits. Operation with iodine specific activity levels greater than the LCO limit is permissible, if the activity levels do not exceed the limits shown in Figure 3.4.16 1, in the applicable specification, for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The safety analysis has concurrent and pre accident iodine spiking levels up to 60.0 pC1/gm DOSE EQUIVALENT I 131.

!' The remainder of the above limit permissible iodine levels shown

. in Figure 3.4.161 are acceptable because of the low probability of a SGTR accident occurring during the established 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months time

(continued) 1 I MARK UP OF NUREG 1431 BASES B 3.4 106 5/15/97

ADDITIONAL INFORMATION COVER SHEET l

ADDITIONAL INFORMATION NO: Q 3.4.1-1 APPLICABILITY: CA, CP, DC, WC REQUEST: Difference 3.4-38 Comment: TSTF-105 has been rejected by the NRC.

FLOG RESPONSE: The July 27,1998 industry traveler status reports indicate the status of TSTF-105 as rejected by the NRC with the TSTF considering. The FLOG has reviewed the traveler and is withdrawing the traveler from the conversion application.

For Diablo Canyon, the CTS will be used which does not require a specific method for !

measuring RCS flow. This difference from the STS is justified by revised JFD 3.4-38.

ATTACHED PAGES: , CTS 3/4.2 - ITS 3.2 , page 3/4 2-13 A, page 9 B, page 6 0 CTS 3/4.4 - ITS 3.4 Enclosure SA Traveler Status Sheet and page 3.4-2 8, page B 3.4-5 A, page 6 8, page 5

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

l

^

POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-1:

a. '

Reactor Coolant System T,g,

b. Pressurizar Pressure, and i:. Reactor Coolant System Total Flow Rate.

a .

APPLICA8ILITY: MODE 1.

4

^

ACTION:

With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5% of-RATED THERMAL POWER within the next hours.

OS*-04-LS-2 (o

SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the parameters of Ta'ble 3.2-2. shall be verified to be within their limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

-f.2.5.2 i'__ __ __..., .

th: : !:ehted ""5 teh! *!= r:te ch !' be deterstnad-to-be-greater-i - . . . . .,--. ..,esem eau.

y_

. ."ri:r b :;: :ti:n '::: 75 ;f "^.TS "E.".".. "0WER-efter-eacir-fuel-1--din;, --i .

t < .,__m ____ ___ ,, ,,_.._._n

.. n. ..... ..... ,,.. .....,__ . . . .

. . . . . . . . . . .. .eys: g,jy.g 4.2.5.3 The RCS loop flow rate indicators s' hall be sub ected to a CHANS CALIBRATION at 1 st once per 18 d=+ renve J-I-Ji- p Ask remye.c{y;kan ** (&,f nths. _

a .t an& g,4j l

4. 5.4 T ACS total flow rate shal b % n d bii:r::i:i:n h::t MI:nc; O*

- : _ c ::t: at least once per 18 months. Within ' d:y :f ;:rf: :f ; th; c5-07fl _

. pr:5 h ier M et bi h nc , the 4 et='---* tier used fer eterminatien-of-steam-rr:re- , fee &rter t---- sture, and fe?@rter venter' ^." 4- the celerimetris E_,_

......,_a......___ u. __,

s.u__..;

. .. _u.._,.3 of-of -tG 1.2.5.5 '5 f::t:ter v::ter* tha!' b *nepected fer feu!i g -d cleaned-as-

- cert: j at !- it :::: ; r 19 centh!. 05-o9. LG 1:

. t

T.'.4 ssisul6tud 16lue of ."CS tGtel IIG rate " hell D: U".;d "in:9-unG M e5-ef 2.5 f;r fl: (including 0.5 f;r f;;ict:r v:nturi f cling}-measurement-

. h;;; 5::: inchd:d '- the deve survef'hnce.- Of-Ol- LCr 1

^

CALLAWAY - UNIT'1 3/4 2-13 Amendment No.15

.gg gje r Ac. @A w LA /g f.~ 2 fr% ATP . of M uw y-r- - ,7 wv w -

~ CHANGE NUPEER LTC DESCRIPTION 04 10 Not applicable to Callaway. See Conversion Comparison Table (Enclosure 38).

1 05 01 LG The designation of how instrument uncertainties are I

treated (nominal, in the analysis, or in the development of the TS limit) is moved to the Bases. The movement of this level of detail out of the specification is consistent with NUREG 1431 and is an example of removing unnecessary details from the TS in accordance with 10 CFR 50.36.

05 02 LS 7 Not applicable to Callaway. See Conversion Comparison Table (Enclosure 38).

05 03 + L.g { c =1 ts t m in E 0-1431. th; m 9 = t te P fe m : [CARM/

~

1 c:' t =: c;,:.:=:CN = th: acS fl= x.t;i; :: 1:=t :=e P 13 vuiha aird tEs FKpali .. .d to i,Gia,li20 th0 Ch.T.Tala are moVeG to Ine pases for Lnc RC3 fl u 10% i

' i- J sip iun 10 4/,[ TeaNa I E Dll'akny. 5ee in li3v.rion Cenvo Sed.UnprTrom~T 3.3.1. Die fgbneforare ~Tk .

05 04 LG sistent w un inaustry traveler 15FF 105, the expilc requi ts that the RCS flow be measured thr use of a precis 1 t balance measurement hat the instrumentation us the perfo ce of the calorimetric flow measur calibrated within a j specified time period perform he measurement is

[

. moved to a lice controlled documen . he requirement to verify the RCS flow is within limits ins withi Technical Specification. This is an exa of r ing unnecessary details from the TS and is acceptab n ed on the auidance provided in 10 CFR 50.36.

~

-4/MT TAA & 2.+./-/

05 05 LG Not applicable to Callaway. See Conversion Comparison Table (Enclosure 38). 1 05 06 LS 8 In accordance with NUREG 1431, if any of the DNB related parameters of pressure, temperature, or RCS flow are found to be outside their limits, the time period required to perform a power reduction would be extended to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The DNB related parameters of Reactor Coolant System (RCS) average temperature, pressurizer pressure, and RCS flow rate are maintained within specified limits in order to ensure consistency with the assumed initial conditions of the accident analyses. The limits placed on the RCS temperature, pressure, and flow ensure that the minimum departure from Nucleate Boiling ratio (DNBR) will be met for each of the transients analyzed. Compliance with the l above limits is verified every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . If a parameter '

DESCRIPTION OF CHANGES TO CURRENT TS 9 5/15/97

( - . - . . - . . . - . . . . - . - . . - . - - - . .

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

INSERT 3A-9 Q 3.4.1-1 CTS SR 4.2.5.4 provides descriptive detail of the method for the determination of RCS total flow rate during a Surveillance. This detailis moved to the ITS SR 3.4.1.4 Bases. These details are not necessary to ensure the RCS total flow rate is within required limits. The requirements l of ITS SR 3.4.1.4 are adequate for ensuring the RCS total flow rate is within required limits.

These detai's are not necessary to be in the TS to ensure the RCS total flow rate is within i

required limits. Moving this information maintains consistency with NUREG-1431. Any change to this descriptive information will be made in accordance with the Bases Control Program described in ITS Section 5.5.14.

i l

CONVERSION COMPARISON TABLE - CURRENT TS 3/4.2 Page 6 of 7 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY 04-06 Actions involving QPTRs exceeding 1.09 would be eliminated Yes No - Actions not in Yes Yes LS-13 in conformance with NUREG-1431. CTS.

04-07 The statement that Specification 3.0.4 does not apply is no Yes No - Exception not Yes Yes A longer needed as revised Actions permit continued operation in CTS.

for unlimited period of time.

04-08 Not used. NA NA NA NA 04 09 Consistent with NUREG-1431. Rev. 1, a Note is added to Yes No - Already in No - Maintaining Yes A permit three OPERABLE excore channels to be used to CTS. CTS wording.

calculate QPTR when one channel is inoperable and power is s 75% RTP.

04-10 The allowed time for the requirement to reset the Power No No Yes No LS-14 Range Neutron Flux - High setpoint during power reduction required by QPTR ACTIONS would be extended to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for Wolf Creek.

05-01 The designation of how instrument uncertainties are treated Yes Yes Yes Yes LG (nominal. in the analysis or in the development of the CTS limit) is moved to the Bases.

05-02 The CPSES specific requirement to verify that the total RCS No Yes No No LS-7 flow is within limits using the plant computer or elbow tap output voltage on a monthly basis is deleted.

05-03 The requirement to p^rfc 2 C!!"r 5L U L? MATIO" :t 1 =t -Ves-

~

f ie "

Ves- A/,- Eas- /\/o-LG -a Fr !" rnth = d the requi.;;;r.t to-normalize the R C J /,,p.ff,y 4 ; g,.,,j,nr 7,f % ve e !

thrch u.c muved to the Bases for th curvei!hne kkNd no-/-fn CTT.

A mwd fn C7T,

_..m____m_ ,__ mu_ nce m. , _ . . _ _ _ _ . _ . , m . ,. .s _

f CN.M/ f, sib k. . .f 05-04 Consist at with ir.dztr; tr:: ler TST" 105. th; nplicit No - Requirement Yes Yes Yes LG em"ir--^^t: that Ita RCS fic; 5: =nur;d t'r;; not in d = precision hat t,alaae .-.;a. ;_.; :n" th ...he g g g f/f , y. g, .fg

~

me r nt : ;

i perfed of perfer 9 g th: re s.. _ .a 4 moved to the Bases.

Lare CONVERSION COMPARISON TABLE - CURRENT TS 5/15/97

IWUSTRY TRAVELERS APPLICABLI TO SECTION 3.4 i

i TRAVELER # STATUS DIFFERENCE # CO M NTS u TSTF 26 Incorporated 3.4 32 - Approved by RC.

TSTF 27. Rev. [ Incorporated 3.4 33 Ap/l yA/4C, -((f jg

y PR Incorporated 3.4 22 Approved by NRC. 7 j L'F 54. Rev. 1 Incorporated NA Ns"Ob3/chE[Enly.72t s4 g 1

TSTF 60 Incorporated 3.4 15 Approved by NRC. Q TSTF 61 Not Incorporated NA Ninor change that is adequately p addressed in the Bases. B TSTF 87. Rev. f Incorporated 3.4 31 A/ /rwe/ yl MAO - 743,49 ;

l TSTF93.Rev.p Incorporated 3.4 17 N M p3r"o M /or E ia'way in OL ,.

i3 4 7,f,4-3 Amendment No. 105. g-?.f x3- A TSTF94;g.ev,l Not Incorporated NA Retained current TS. g 2.4tf e m. . . . , a f_. . ,

.z m - .

v T. ..

g.,-

" "v g.g+/-i g3r. m/ k

~

TSTF 108hev.1 Not Incorporated NA LCO 3.4.19 does not apply.

TSTF 113. Rev. [Yncorporated 3.4 39 d 7,4,/(,-/[g241;//-3 -S 2,' - 3//

TSTF 114 Incorporated NA Approved by NRC: Bases 3.4.7 l

t changes only, TSTF 116. Rev. [Yncorporated 3.4 36 [ 74/3-2 fg 7 y A/l h TSTF 136 Incorporated NA BEIN #ald U .N changes -

on1y. ,, -rg y.4g}

TSTF 137 Incorporated NA BMMNch7(($nly. ng_y,4 g '

TSTF 138 Not Incorporated NA Inconsistent with RCS loops requirements of ITS 3.4.5 and ~

i 3.4.6.

b l

TSTF 151gev. / Incorporated NA Bases 3.4.11 changes only.WS.6pf f TSTF 153 Incorporated 3.4 01 ,A//*8 V M k N A C - 74-J,f#1 TSTF 162 y Incorporated NA Ms 3N.h[chInhI>'only. fat-34.-g; Incorporated g g . 3.4 45, See also hs3'.i-[and 3.b0.-

urn-aav Incorporated 3.4 35 j l

g y,4//-p p f.2 7j_

% d D5N- Incorporated 3.4 10 DCPP only./)pr,wllyN#d.7;qi y,4 WOG 87, ge v. 2. Incorporated 3.4 47 f)7.f;//-f 7,f "f,ee I MARX UP OF WOG STS REV 1 (NUREG 1431) 5/15/97

'RCS Pressure Terperature, and Flow DNB Liaits 3.4.1 1

1 SURVEILLANCE REQUIREENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 Verify pressurizer pressure is a E209 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months e B PS psig.

I SR 3.4.1.2 Verify RCS average temperature is s: 5et 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months - 8 PS .

F.

i' SR 3.4.1.3. Verify RCS total flow rate is a 984;409 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months gpe. e 8 PS t .

1 1

. I

! SR 3.4.1.4 ... ... ......... . N0TE -

Not required to be performed until 24 Mur:. 2:3 4 40F i' after a 90 g

RTP.

............................................. 3,j ]

,, e da /

.; Verif$,y i,7;;;;ier, M;t Mier.;; tMt 3 months 384 ui

. RCS tc tal flow rate is a 284:409 - -

gpe, ;B PSj by frecisten ku+ lalance +Q l

4

)

MARK UP OF WOG STS REV 1 (NUREG 1431) 3.4 2 5/15/97 5

RCS Pressure Temperature, and Flow DfB Liaits B 3.4.1 BASES ,

SURVEILLANCE SR 3.4.1.4 ,

,j g, REQUIREENTS Q ,4 J,,f,,,,

(continued) Measurement of RCS total flow rat y p,...,. . a a precisier, $ 3,4 /-/

1eriatric k;t beler.,. once every Ignonths a -

. allows the installed RCS flow instrumentation to be ceHbreted M and verifies the actual RCS flow rate is greater than or equal to the minimum required RCS flow rate.

& 2.4./-l

re,. =r e i ,q '. . +- ,. d. .

The Frequency of 3 months reflects the importance of verifying flow after a refueling outage when the core has been altered, which may have caused an alteration of flow resistance.

8 This SR is modified by a Note that allows entry into MODE 1, without having performed the SR. .and placement of the unit in the best condition for performing the SR. The Note states that the SR is not required to be performed until 24-hotes-7mfter a 995 Et RTP. This exception is appropriate since the heat balance requires the plant to be at a minimias of 996-93 RTP to obtain the stated RCS flow accuracies : ." ,

- The Surveillance shall be performed within M .tra M after reaching 994-93 RTP.

REFERENCES 1. FSAR, 0;; tier. 15 w -

i m

MARK UP OF NUREG 1431 BASES B 3.4 5 5/15/97

CHANGE MHEA JUSTIFICATION but would limit the exception to prior to entering MODE 2. This change is consistent with traveler C-C^. ; TJT/~-aff, $ f, f,//-g 3.4 36 SR 3.4.13.1 and ACTIONS for LC0 3.4.15 are revised with the addition of a note per traveler TSTF 116,"r;. 1. The note addresses the concern tQ f.f,/3-2 that an RCS water inventory balance cannot be meaningfully performed unless the unit is operating at or near steady state conditions. The Note added to the surveillance provides an exception for operation at less than steady state conditions. The RCS water inventory balance will only be allowed to be deferred for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after re establishing steady state conditions.

3.4 37 The primary to. secondary leakage limits are revised per Callaway OL Amendment No.116 dated October 1,1996.

3.4 38 m tent with traveler im-105, the details on nuc Dy which is verified are moved f the RCS 3.4.1.4 to the Bases.

Moving this informatlo ows the use of precision heat balances, elbow taps, r acc thods in order to perform this verificat nd is consistent with the -

ev. 1 philo moving clarifying information and descriptive s_out ha TS to the Bases s ,.y /, f, & 9, f,, ' f) Y,f,/- /

0 rwer.cfon % w iron' g Tibla Ene/sr m '

3.4 39 The shutdown regtfirements of ITS .4.11wouldre$$qui e the plant to reduce T.,, to < 500*F within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , rather than go to MODE 4, to address the concern of entering [COMS] LC0 3.4.12 Applicability with inoperable PORVs. For consistency, the shutdown requirements of ITS 3.4.16 are a revised to allow 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to reduce T _ to < 500*F.

This change is c sistent with traveler TSTF-1136ZA/JFEgf

" gA-f] 8IA //~S TM-M2 3.4 40 Consistent with traveler 499-99?'the Note to SR 3.4.1.4 would be & S f. / . 2.

modified to provide additional ~ time to perform an RCS precision flow rate measurement. The time allowed would be changed from 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 7 days. This change is acceptable because other indication of RCS flow is available (SR 3.4.1.3, RCS total flow meters) and additional time normally would be required to establish plant conditions suitable for the precision heat balance. Since this parameter does not normally change significantly and the flow meters can be used in the interim, there'is no need to perform this SR within the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period specified in NUREG 1431 Rev. 1. The 7 day period provides sufficient time to establish steady state plant thermohydraulic conditions and obtain equilibrium xenon. In addition, the THERMAL POWER specified in the Note would be changed from the generic value in brackets (90% RTP) to 95% RTP. This change is acceptable because it specifies a power level in better agreement with current operating procedures for performing a precision heat balance. Current TS do not specify a power level for

-e this measurement.

3.4 41 Not applicable to Callaway. See Conversion Comparison Taole (Enclosure 68).

JUSTIFICATION FOR DIFFERENCES TS 6 5/15/97

8 .

l j ECT ON 3 4

Ve r re$,fne ed Err n$Nr.4 -led.

DIFFERENCE FROM NUREG-1431 APPLICABILITY Nt# EEL DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY 3.4-37f The primary to secondary leakage limits are revised per No No No Yes Callaway OL Amendment No. 116 dated October 1. r36.

/

3.4-38 r-W+rt iti. tie Ai ";710";. um e6.h r. tk -Vee- -Ves- -Ves- -Ves-n 2 .t.h Cl N T.T'" "' " " " }'*' % M M$uH 3.4-39 The shutdown requirements of ITS 3.4.11 would require the Yes Yes Yes Yes plant to reduct T ,to < 500*F within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , rather than 90 to HDDE 4 to address the concern of entering [COMS]

LCO 3.4.12 Apot .cability with inoperable PORVs. For consistency + e shutdown requirements of ITS 3.4.16 are also revised to allow 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to reduce T, to < 500*F.

Tis change is consistent with traveler TSTF-113.

3.4 40 The Note to SR 3.4.1.4 would be modified to specify a No - See CN 3.4-51. No - See CN 3.4-34. Yes Yes plant-specific reactor power and to provide additional time to perform an RCS precision flow rate measurement.

3.4-41 LC0 3.4.1 is revised to reference Tables 3.4.1-1 and Yes - Allowance No No No 3.4.1 2 for vS total flow rate limits for DCPP Units 1 added per Amendment and 2 resy 'ively. 60/59.

3.4-42 An exception to SR 3.4.14.1 frequency to le3k test Yes No No No PIVs 8802A 8802B and 8703 has been added. This change is consistent with the DCPP current TS.

3.4 43 A new Condition C is added to LCO 3.4.1 to reflect the No No Yes No current TS of Wolf Creek for RCS flow rate.

3.4 44 Steam generator levels for MODES 3. 4. and 5 are specified No No

-No-[er Yes g g*g:.7 to ensure SG tubes are covered. The O llary current TS did not ensure tube coverage.

t CONVERSION COMPARISON TABLE - NUREG-1431 5/15/97

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.1-2 APPLICABILITY: CA, CP, DC, WC REQUEST: Difference 3.4-40 Comment: WOG-99 has not yet become a TSTF.

FLOG RESPONSE: WOG-99 has been designated TSTF-282 which is currently under NRC review. No changes to the ITS mark-ups were made in the process of assigning this traveler a TSTF number. As explained in Enclosure 6B to Attachment 10, JFD 3.4-40 does not apply to CPSES or DCPP. Those plants are retaining their CTS, as explained under JFDs 3.4-34 and 3.4-51, respectively. Callaway and Wolf Creek continue to pursue the changes proposed by this traveler.

ATTACHED PAGES: , CTS 3/4.2 - ITS 3.2 A, page 10 0, CTS 3/4.4 - ITS 3.4 Enclosure SA, Traveler Status Sheet A, page 6 l

i i

l CHANGE NUMBER HSE DESCRIPTION is found to be outside the required limit. 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months are allowed in order to restore the parameter to within the

limit. If the parameter is not restored to compliance within the required time. the plant must be shut down.

The revised completion time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is acceptable to l allow transition to the required plant conditions in an orderly manner without unnecessarily initiating any undue plant transients and on the small likelihood of a severe event occurring during the extended time period.

05 07 M This surveillance for measuring RCS flow by precision heat balance is modified to add a footnote that corresponds' to the Note for ITS 3.4.1.4. The footnote requires that the surveillance be performed within 7 days of achieving 95% RTP. This is more ress % v've in that it ties the surveillance to the beginnig of a cycle. This is acceptable because other indication of RCS flow is available (RCS flow meters) and time is provided to establish plant conditions suitable for the precision heat balance. This is consistent with traveler L'^C 00. In addition. the THERMAL POWER specified in footnote would be changed from the generic value p ovided in NUREG 1431 to a plant specific vibe of a 95% RTP. This I

change is acceptable because it specifie a THERMAL POWER in better agreement with current operati g procedures for performing a precision heat balance. C rrent TS do not specify a power level for this measur nt.

73 rf:~ 2.12 O 3' U4 !

05 08 Not used.

05 09 LG The requirements for inspecting and cleaning the feedwater flow venturi would be moved to licensee controlled documents. These details are not contained in NUREG 1431.

This is an example of moving unnecessary detailed information from the TS and is acceptable.

05-10 A The requirement to verify RCS flow rate within limits prior to operation above 75% RTP after each fuel loading and at least every 31 EFPDs would be eliminated from the SRs for DNB parameters. This is acceptable based on the l

requirement to verify RCS total flow rate once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months in accordance with SR 3.4.1.3. In addition. F",

is verified within limits prior to operation above 75% RTP after each fuel loading and every 31 dag . The LCOs.

Actions and SRs for F", and RCS total flow rate are

! adequate to address any DNB concerns related to these parameters.

DESCRIPTION OF CHANGES TO CURRENT TS 10 5/15/97

TRAVELER # STATUS DIFFERENCE # COMENTS N Incorporated 3.4 40 Applicable to Callaway and Wolf l

M,ap3. Creek only. 4 2 f. /-2 -,.n- ~. . , .,,

' 9_ff ~ Incorporated 3.4 49 $ 7.f./.2-/-g .:'. f ;;7 1

l l

l I

1 l

I i

MARK UP OF WOG STS REV 1 (NUREG 1431) 5/15/97

CHANGE NUPEER JUSTIFICATION

. b:

4 but would limit the exception to prior to entering MODE 2. This change is consistent with traveler = 00.- TJTF-aff, d) f. f.//-g 3.4 36 SR 3.4.13.1 and ACTIONS for LC0 3.4.15 are revised with the addition of l a note per traveler TSTF 116,"re,1. The note addresses the concern $ f.f,/3-2 that an RCS water inventory balance cannot be meaningfully performed unless the unit is operating at or near steady _ state conditions. The Note added to the surveillance provides an exception for operation at less than steady state conditions. The RCS water inventory balance will only be allowed to be deferred for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after re establishing steady state conditions.

3.4 37 The primary to secondary leakage limits are revised per Callaway OL Amendment No.116 dated October 1,19%.

i 3.4 38 tent with traveler TSTF 105, the details on iva Dy which 1

the RCS is verified are moved f 3.4.1.4 to the Bases.

Moving this informatlo ows the use of precision heat

. balances, elbow taps, r acc thods in order to perform j this verificat is consistent with the -

ev. 1 i

philo moving clarifying information and descriptive s_out TS to the Bases / y , f, , '$ .7 f, /-/

Onver.rton Cwri.ron~y , jfy/, f C,j$

734/e Enc /s.run 3.4 39 The shutdown regtfirements of ITS .4.11 would requi e the plant to reduce Tm to < 500*F within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , rather than go to MODE 4. to address the concern of entering [COMS] LC0 3.4.12 Applicability with inoperable PORVs. For consistency, the shutdown requirements of ITS 3.4.16 are al revised to allow 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to Aede r T._ to < 500 *F.

This change is c sistent with traveler TSTF~ 113/.rA/J2Egt gA-f] 8 M //'d

^

~

7J1r--M2 3.4 40 Consistent with traveler 4GIHMh#the Note to SR 3.4.1.4 would be & S f. /-2.

modified to provide additional time to perform an RCS precision flow rate measurement. The time allowed would be changed from 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 7 days. This change is acceptable because other indication of RCS flow is available (SR 3.4.1.3, RCS total flow meters) and additional time normally would be required to establish plant conditions suitable for the precision heat balance. Since this parameter does not normally change significantly and the flow meters can be used in the interim, there'is no need to perform this SR within the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period specified in NUREG 1431 Rev. 1. The 7 day period provides sufficient time to establish steady state plant thermohydraulic conditions and obtain equilibrium xenon. In addition, the THERMAL POWER specified in the Note would be changed from the generic value in brackets (90% RTP) to 95% RTP. This change is acceptable because it specifies a power level in better agreement with current operating procedures for performing a precision heat balance. Current TS do not specify a power level for this measurement.

3.4-41 Not applicable to Callaway. See Conversion Comparison Table (Enclosure 6B).

JUSTIFICATION FOR DIFFERENCES TS 6 5/15/97

i 1

ADDITIONAL INFORMATION COVER SHEET l

ADDITIONAL INFORMATION NO: O 3.4.2-1 APPLICABILITY: CA, CP, DC, WC 1

REQUEST: Difference 3.4-33 l

l Comment: TSTF-27 Rev. 3 is still pending NRC approval.

FLOG RESPONSE: The July 27,1998 industry traveler status reports indicate the status of TSTF-27, Rev. 3 as approved by the NRC. The proposed wording in TSTF-27, Rev. 3 was i

modified from TSTF-27, Rev. 2, and these modifications have been incorporated into the ITS.

l The FLOG continues to pursue the changes approved in TSTF-27, Rev. 3.

i ATTACHED PAGES:

l Attachment 10, CTS 3/4.4 - ITS 3.4 Enclosure SA, Traveler Status Sheet Enclosure 58, page B 3.4-8 Enclosure 6A, page 5 i

l l

IP00STRY TRAVELERS' APPLICABlf TO SECTION 3.4

, ,- . TRAVELER # STATUS DIFFERENCE # C0fMENTS '

_TSTF 26 Incor ppo _

3.4 32 Approved by IRC. _

TSTF 27, Rev. [ Incorporated 3.4 33 Ajpmell y mad.- [ I'd g !

1422

~

Incorporated Approved by NRC.

, [

TSTF 54, Rev.1 Incorporated NA S$s~Mb37cdnY5nly.7pg4 y9

.L TSTF 60 Incorporated 3.4 15 Approved by NRC.

TSTF 61 Not Incorporated NA Minor change that is adequately 5 y addrested in the Bases. 1 TSTF 87, Rev. f Incorporated 3.4 31 Aff8rwe/ly NA6- 7g-3,44 U TSTF 93, Rev.p Incorporated 3.4 17 M p3r"oi M /or N a'way in OL ',.

i., 3 4 7,f,4-3 Amendment No. 105. g -?.$ g; 'r:

TSTF94;/ev,1 Not Incorporated NA Retained current TS. rg y,4( L k

= w, f=. ! :..:.. n .w s ..;. 2 gup-i g _m/ a TSTF 108. Rev.1 Not Incorporated NA LCO 3.4.19 does not apply.

TSTF 113. Rev. [Yncorporated 3.4 39 8 74,/4-/ $24//-3 S ;',N//

TSTF 114 Incorporated NA Approved by NRC: Bases 3.4.7 changes only.

TSTF116.Rev.[Yncorporated 3.4 36 [;TA/2-2 g7,y ofj TSTF 136 Incorporated NA B(MI43 #aM N .9 changes only. ,, Tg-fu9 TSTF 137 Incorporated NA B[MTINch7[3e $nly. q_gq 9, TSTF 138 Not Incorporated NA Inconsistent with RCS loops requirements of ITS 3.4.5 and 3.4.6.

4 TSTF 151gev. / Incorporated NA Bases 3.4.11 changes only. fk-S.hpf [j ,

TSTF 153 Incorporated 3.4 01 A //* V M k A/A C 5,4#1 TSTF 162 y Incorporated NA Es 37.9[cEnhI>'only. 72L-34-g ,,

Incorporated g Q 23, 3.4 45, See also Ns3'.i-[and 33I20. ,

M Incorporated 3.4 35 g jag-p -co f.f , fj_ [

m -MbNN- Incorporated 3.4 10 DCPP only./)pipw//yN#d,gg, WOG 87, ge v. 2. Incorporated 3.4 47 $ 7f:// 7,0 -'.f.ee MARK UP OF WOG STS REV 1 (NUREG 1431) 5/15/97

4

! RCS Minista Temperature for Criticality 4

8 3.4.2 j _ m.,

l BASES l APPLICABILITY necessary to allow RCS loop average temperatures to fall below i j (continued) T,,w. which may cause RCS loop average temperatures to fall

below the temperature limit of this LCO.

l 1 '

1 l ACTIONS Al ,jj y,} ;

1

, If the pal aseters that are outside the limit cannot be restored, ;

l the plant must be brought to a MODE in which the LCO does not l apply. achieve this status, the plant must be brought to '

HOBE-G .EEmmas within 30 minutes. Rapidreactor8.24: 6en-/

shutdown can be readily and practically achieved within a ,

a 30 minute period. The allowed time is reasonable, based on 1 j -

operating experience, to reach MODE 3 in an orderly manner and

without challenging plant systems.

i

SURVEILLANCE SR 3.4.2.1 !

l REQUIREMENTS l

! RCS loop average temperature ired to be verified at or l i above 541- 'F every j '.= i= T.,, :1:= =t =:d .y. ;._ PT_* _n : M7"I.tind::

1::- T.,, 2. : T,,,-L ivietier.. $ 7,f.6 )

] ; . ;Ot; ;,,:1fi;; ;l.; ';". 2.;r. ; 6 ."C'; leei, ;ver. ; to ,;. tur;

4 1: '__517'F ;-d,___..____m__

_ , _ ,__ o,,,

tb T ,, T, i;hti=. 1= ,i= T,,,

___m..___ _ _ . .

-lar- i;

, _ , , < _ , _ . m_ , , , ,

l

....... ,, .,,, . ,,, ..... , .,,,........ - .2 . . . . . - . . -

i r;,;;ir;. r.t W.: t ;ditir,r.;.1 ;.cr.in;. The SR to verifv l RCS loop %erage temperatures eveM 30-efnutes-

! - 1; 77:::_ ' r" te ; :::d tb i=i;td ;thti= Of tb S 24:;2-/

l w

conHnuouff a not i.r conc tr+<n+

l svi-/-L o-lhee enHne J'unveillancor wArc), are -lyricallv o 2 ort dre j'erkvmed oneef 15.. 0 er rArf+. Arned pe

.rn ooVr/fon,h le .rener#ve REFERENCES 1. FSAR, Secti;n r .

j 'Is AC E 4xn,feyaIvre duri

~

afj orn a c4 -h cer/ica ffh aJ u,,')/ en.rure -/-J,}- +4e l l {

, nint- .

+en,.a~ h g p,,. c,, -yc_, t,- r, n aJ. <r

Ye < lE fr */ p sac /esl. _ R . * !!n X

MARK UP OF NUREG 1431 BASES B 3.4 8 5/15/97

CHANGE NUMBER JUSTIFICATION l

F'- 3.4 30 An LC0 3.0.4 exception is added to the Actions for LC0 3.4.12. This is consistent with the current licensing basis. Additionally, circumstances could arise where increasing MODE would reduce the risk l of a low temperature overpressurization event. In these cases it would I be unwise to maintain the plant in a lower MODE configuration. l Increasing plant MODE may also be the expedient way to exit a low temperature overpressurization potential when operating within a Condition. This option should be retained as exists in the current Technical Specifications.

3.4 31 These ACTIONS in ITS 3.4.5 and 3.4.9 are modified to reflect their LCO.

The position of the reactor trip breakers and the power supply status of the CRDHs are not LCO requirements; therefore, the CONDITIONS and ACTIONS are revis' . As worded in NUREG 1431 Rev. 1, these ACTIONS could prect t.ertain testing in MODE 3. A more generic action, which assures rods can not be withdrawn, replaces the specific method of precluding rod withdrawal. The specific methods are added as examples to the Bases. The revised ACTIONS still assure rod withdrawal is precluded and this detail is not required to be in the TS to provide l adequate protection of the public health and safety. No technical changes result from this change. These changes are consistent with traveler TSTF 87, k;.1. ~7T-74-#4f-1 3.4 32 In accordance with industry traveler TSTF 25, the ACTION would be changed to specify taking the plant to a MODE for which the LCO is not applicable. This change maintains the consistency between the Mode of Applicability and the Required Action which requires the $4 ode of Applicability to be exited. .

1 3.4 33 The Frequency of SR 3.4.2.1 to verify operating RCS loop average temperature at or above [551]*F is changed to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months from the current surveillance frequency of 30 minutes. . The SR to verify operating loop average temperatures every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months de e""4 dent', '. :gert t: prr;; ,t -

air.;i;:rt:d 12ttr. sf tlu LC0 ;r.iconsiders indications and alarms that are continuously available to the operator in the control room. O ~I This change is based on industry traveler TSTF 27. -te. 2. Il 7.I4 /J 3.4 34 Not applicable to Callaway. See Conversion Comparison Table (Enclosure 68).

3.4 35 This change adds a Note to SR 3.4.11.1 and SR 3.4.11.2 stating that the SRs are only required to be performed in MODES 1 and 2. The Actions Note, "LC0 3.0.4 is not applicable." is intended to allow MODE changes for testing purposes (per the Bases). This allowance is properly presented as an SR Note. A properly placed exception (i.e., an SR Note exception) would not allow the SR to be considered to be met until the appropriate conditions were available for it to be perfomed without entering the Actions. The Note to these SRs would allow startup in MODE 3 if the SR had not been performed during the required frequency, JUSTIFICATION FOR DIFFERENCES TS 5 5/15/97

l ADDITIONAL INFORMATION COVER SHEET 4

ADDITIONAL INFORMATION NO: Q 3.4.3-1 APPLICABILITY: CA, CP, DC, WC REQUEST: ITS 3 4.3 Bases References Comment: WCAP-14040-NP-A, Rev. 2 January 1996, has replaced WCAP-7924-A, April 1975. Please summarize the differences / applicability to the FLOG.

FLOG RESPONSE: WCAP-14040-NP-A, Rev.1 was NRC approved as an acceptable reference for " Methodology Used to Develop Cold Overprossure Mitigating System Setpoints and RCS Heatup and Cooldown Curves" by SER dated 10/16/95 with minor comments which did not affect the SER. These comments were incorporated and the WCAP-14040-NP-A was issued as Revision 2 in January 1996. NRC acceptance of this WCAP as a reference was based upon the following key elements:

1) The WCAP incorporates state of the art fast neutron radiation transport.

2) The WCAP cold overpressure mitigating system satisfies SRP Section 5.2.2 and BTP RSB 5-2.

3) The WCAP fracture mechanics calculation conforms to 10CFR50, Appendix G and SRP Section 5.3.2.

4) The WCAP conforms to Reg. Guide 1.99, Rev. 2 in calculation of the adjusted reference temperature.

5) The WCAP conforms to 10CFR50, Appendix G for methodology for calculating minimum temperature in the P-T limit curves.

6) The WCAP satisfies the provisions of the draft generic letter published in the Federal Registar for comment of June 2,1995.

These items are consister.t with the STS reviewer's Note on STS 5.6.6.

Plant Specific Discussion:

This reference was accepted and incorporated in the CTS by OL Amendment No.124 dated April 2,1998.

ATTACHED PAGES:

None

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.4.4-1 APPLICABILITY: CA, CP, DC, WC REQUEST: ITS 3.4.4 Bases Comment: The Bases refer to the DNDR limit in the safety limits. Where is it? (this appears to be a problem with the STS, as well as these conversions).

FLOG RESPONSE: As described in the Applicable Safety Analyses Bases for ITS Section 2.1.1, the DNBR limit is: "There must be at least 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB." The actual numerical value is specific to a given DNBR correlation and analytical methodology. The correlations and methodologies are NRC-approved. More than one correlation or methodology, as generally documented in the FSAR, may be used depending on core design and the particular transient being analyzed. For this reason, a more general term such as "DNBR limit" is used. This convention has been used throughout the Bases for ITS Sections 2.0,3.1,3.2, 3.3, and elsewhere in 3.4.

In the process of responding to this RAl, it was noted that all FLOG plants except DCPP and :

CPSES have a markup methodology error in the second to last paragraph of the Applicable Safety Analyses Bases for ITS Section 3.4.4. The acronym "SL" should have been struck-through; this is addressed under Comment Number 3.4. Gen-1.

Plant Specific Discussion The Background Bases for ITS Section 2.1.1 provides additional discussion of the DNBR corre,lations and methodologies used for Callaway. In addition, references to pertinent FSAR sections are given.

ATTACHED PAGES:

None l

- ADDITIONAL INFORMATION COVER SHEET !

l ADDITIONAL INFORMATION NO: Q 3.4.5-1 APPLICABILITY: CA, WC REQUEST: Change 1-14 LS-22. (Callaway and Wolf Creek)

Comment: The change d;scussion is not adequate. The NSHC contains the necessary ' 4 justification.

i FLOG RESPONSE: DOC 1-14-LS-22 is revised to read:

I "The LCO and ACTION b of Specification 3.4.1.2, " Reactor Coolant System, Hot Standby,'

would be revised to require that two reactor coolant loops be OPERABLE. Loop operation :

requirements would also be revised to be contingent on Rod Control System status. The j requirement to have a third OPERABLE reactor coolant loop would be deleted, consistent with ,

- NUREG-1431. This is acceptable because the MODE 3 decay heat removal requirements are 'l

- sufficiently low that a single RCS loop with one RCP running is adequate to remove core )

decay heat. A second RCS loop ensures redundant capability for decay heat removal. When !

the Rod Control System is capable of rod withdrawal, two loops must be in operation to ensure !

accident analysis assumptions are satisfied. When rod withdrawal is precluded, only one loop )

is required to be in operation to satisfy MODE 3 accident analyses. The MODE 3 accident j analyses which assume only two RCS loops in operation include the Uncontrolled RCCA Bank i Withdrawal from Subcritical and the hot zero power RCCA ejection events. The initial l conditions and analysis assumptions for these events will be unchanged since two loops must 1 still be in operation during MODE 3 when the Rod Control System is capable of rod withdrawal. I These reactivity transients rely on the Nuclear Instrumentation System's high flux trips for event termination which occurs very rapidly (on the order of seconds). There would be no benefit of having a third RCS loop OPERABLE for these transients since by tlie time the loop could be brought into operation, the event would be over for all practical purposes."

ATTACHED PAGES:

- Attachment 10, CTS 3/4.4 - ITS 3.4 Enclosure 3A, page 3 I

I

k CHANGE NLPEER NSBC DESCRIPTION wa 1 12 M The Actions are changed to separate the required actions for only one required RHR loop OPERABLE and no required RHR loops OPERABLE. These revised Actions are consistent with the Actions which are required under this LCO in <

NUREG 1431 Rev. 1, and are more conservative than current required Actions.

1-13 M An additional restriction is placed on the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months allowance to remove all RHR pumps from operation. This restriction precludes further draining of the RCS which is an acceptable, conservative action. This change is in conformance with NUREG-1431 Rev. 1.

1 14 .LS 22 and Action b are revised to require oops OPERABLE w1 s in opera n the rod control system is capable of r 1 and one loop in operation whe od control syst a able of rod al. This change is consistent with 3

. 1. ZWStT~ 2h-3A $ 3.+.6~~l 1 15 H A steam generator (SG) level corresponding to 10% of the wide range does not cover all of the SG ttbes. To qualify as a valid heat sink, the tubes must be covered. This is -

a more restrictive change. [Four percent of the narrow range span is specified at the higher temperatures of MODES 3 and 4 whereas 674 of the wide range span is 6 5. f,5~--2 specified for MODE 5. th values ensure SG tubes are covered. The E0Ps cite he 4% narrow range level to I ensure heat sink adequ y.]

44 %

1 16 A Consistent with the intent of traveler TSTF 153, this change revises the note that permits up to I hour "deenergization" of RCP/RHR pumps. The revised wording clarifies the intent of the note to allow the pumps to be

" removed from operation" instead of "deenergized", thus permitting other means of removing the pumps from service.

With this change the pumps are not reauired to be deenergized to use the note (e.g., the pumps may be

~

isolated, etc.). The change is considered to be administrative because from the standpoint of providing an exception to the LC0 requirements (to maintain the operability and operation of the pumps), the revised wording is equivalent.

. M 1-17 - Not applicable to Callaway. See Conversion Comparison Table (Enclosure 38).

1 18 Not used.

Di.SCRIPTION OF CHANGES TO CURRENT TS 3 5/15/97

INSERT 3A-3A Q 3.4.5-1 The LCO and ACTION b of Specification 3.4.1.2, " Reactor Coolant System, Hot Standby,'

would be revised to require that two reactor coolant loops be OPERABLE. Loop operation requirements would also be revised to be contingent on Rod Control System status. The requirement to have a third OPERABLE reactor coolant loop would be deleted, consistent with NUREG-1431. This is acceptable because the MODE 3 decay heat removal requirements are sufficiently low that a single RCS loop with one RCP running is adequate to remove core decay heat. A second RCS loop ensures redundant capability for decay heat removal. When the Rod Control System is capable of rod withdrawal, two loops must be in operation to ensure accident analysis assumptions are satisfied. When rod withdrawal is precluded, only one loop is required to be in operation to satisfy MODE 3 accident analyses. The MODE 3 accident analyses which assume only two RCS loops in operation include the Uncontrolled RCCA Bank Withdrawal from Subcritical and the hot zero power RCCA ejection events. The initial conditions and analysis assumptions for these events will be unchanged since two loops must still be in operation during MODE 3 when the Rod Control System is capable of rod withdrawal.

These reactivity transients rely on the Nuclear Instrumentation System's high flux trips for event termination which cccurs very rapidly (on the order of seconds). There would be no benefit of having a third RCS loop OPERABLE for these transients since by the time the loop could be brought into operation, the event would be over for all practical purposes.

i

)

d a

...g. -

w- -

m w..m. .-r-4 e a ....... -_. , _ _ -

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

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.5-2 APPLICABILITY: CA, WC REQUEST: ITS SR 3.4.5.2 (also SR 3.4.6.2 and SR 3.4.7.2) (Callaway)

Change 1-15M l Comment: The sections of the ITS use the phrase "or equivalent" yet the term is not explained in the change or in the ITS Bases. According to the information provided narrow range level is used at the higher temperatures (Modes 3 and 4) and wide range level is used at the lower temperatures (Mode 5). If "or equivalent" means using the wide range at higher temperatures and the narrow range at lower temperatures are the levels specified applicable at the different temperatures? If not, anM are the equivalent ,

leveIs to the values specified in the ITS and how were they determined?

FLOG RESPONSE: At Callaway, the top of the highest steam generator (SG) tube is 344 4

inches above the tube sheet. The wide range instrumentation provides level indication from 7

, inches above the tube sheet (0% indication) to the moisture separators (a range of 559 inches). The narrow range instrumentation provides level indication between 438 and 566 inches above the tube sheet for its 0-100% indication (the use of a uommon upper tap results in 100% level indication on both wide range and narrow range ne tnally being at the same l 566 inches above the tube sheet). A calculation was performer :orrelate the top of the l highest tube to the wide range scale for MODE 5 conditions (M wide range instrumentation is I calibrated for cold conditions), with margins added in for instrument loop errors and readability, I resulting in the specified 67% wide range level. A minor error in the calculation was corrected, resulting in the specified 66% wide range level value cited in the attached pages for Callaway.

Since the zero reference for the narrow range level instrumentation is nominally 96 inches i above the top of the highest tube, the 4% value specified for MODES 3 and 4 was chosen ;

since it is used throughout the EOPs for heat sink indication and is familiar to the operators. In l the main control room there is one Class 1E wide range level indicator per SG and there are l four narrow range levelindicators per SG, of which three per SG are Class 1E. The "or equivalent" phrase would allow the use of wide range level instrumentation in MODES 3 and 4 in the unlikely event all narrow range level instrumentation were unavailable for a required SG; in MODE 3 this unlikely scenario would result in ITS LCO 3.3.3 non-compliance and would invoke Required Action (s) under PAM Instrumentation. Conversely, the "or equivalent" phrase would allow the use of narrow range level instrumentation in MODE 5 if the one wide range level indicator per SG were unavailable. This flexibility is similar to the approach under which Vogtle was licensed wherein their MODES 3-5 RCS specifications required SG water level to be above the highest point of the SG U-tubes. We are specifying water levels that ensure the same, yet allow the use of all available instrumentation. Before the "or equivalent" instrumentation were used in a given MODE, process measurement effects on the altemate instrument's calibrated span would be considered. Due to the unlikely event of either scenario presenting operational limitations, given the reduced RCS loop requirements in MODES 3-5

,and the instrumentation redundancy, we do not see the need for a pre-determined correlation

. between the wide range ano narrow range level indications; however, we reserve the right to exercise that option should the need arise.

Wolf Creek reviewed this particular comment for applicability to Wolf Creek and concurs with l the use of the phrase "or equivalent" in the ITS and ITS Bases. Wolf Creek believes that it is appropriate to change their plant-specific value to 6% narrow range (including uncertainties)

l' l

l since it is used throughout the Emergency Operating Procedures (EMGs), it has operator l awareness because of the EMG familiarity, and ensures an SG water level approximately 100 l inches above the top of the highest SG tube. Wolf Creek has done a review of the drawings and design documents and has determined that for MODE 5 conditions (the wide range instrumentation is calibrated for cold conditions),66% wide range level corresponds to the top

, of the highest tube, with margins added in for instrument loop errors and readability. The need j for flexibility to use either narrow range or wide range indication is most evident when placing l the SGs in wet layup conditions. The narrow range instruments are "jumpered" to indicate a l constant 50% level. This precludes a feedwater isolation signal at approximately 78%. The l operators use SG wide range indication to maintain and monitor SG level. Additionally, the

narrow range instruments are calibrated for normal operating pressure and temperature conditions while the wide range instruments are calibrated for shutdown conditions.

The Callaway and Wolf Creek ITS Bases have been modified to explain the "or equivalent" l phrase.

ATTACHED PAGES:

1 Attachment 10, CTS 3/4.4 - ITS 3.4 l Enclosure 2, page 3/4 4-5 Enclosure 3A, page 3 Enclosure SA, pages 3.4-12 and 3.4-13 Enclosure SB, pages B 3.4-26, B 3.4-32, B 3.4-33, B 3.4-34, B 3.4-36, and B 3.4-37 j Enclosure 6A, page 7 '

l Enclosure 68, pagn 5 -

l 1

i l

l 1

4 1

I

REACTOR COOLANT SYSTEM

..d,,...

}

u ,

COLD SHUTDOWN - LOOPS FILLED

' " LIMIlING CON 01I10N FOR OPERAT10N

3. 4.1. 4.1 At least one residual heat removal (RHR) loop shall be OPERABLE anc /-4 7-Z.5 in operation and either:

a. One additional RHR loop shall be OPERABLE #, or

b. The secondary side - ' vel,,of at least two g e,a g-rators /-/S-/y shall be greater han-WWrjof the wide range.~

la'nt Ys f ed##.

NM APPLICABILITY: MODE 5 wi h rea o N #- A- -

ACTION:

a. With one of the RHR loops inoperable and with less than the required steam generator level, immediately initiate corrective action to return the inoperable RHR loop to OPERABLE status or restore the required team generator level as soon as possible.

reguiredWithKHf igenMe or. /-/4-/f -

b. Yno RH . loop in operation, suspend all operations involving a -

reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop tojoperation.

blWlf JhfMf Y

.- SURVEILLANCE REOUIREMENTS

4. 4.1. 4.1.1 The secondary side water level of at least two steam generators when required shall be determined to be within limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.1.4.1.2 At least one RHR loop shall be determined to be in operation 4ad--

/-4[-/ 6-rc ct: ;d er.: at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

( ' cOSEN

""' ti r; b gj,,,

0ne RHR loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other RHR loop is OPERABLE and in operation.

1. A reactor coolant pump shall not be starte g unless the secondary water temperature of each steam generator is less than 50*F above each of the Reactor Coolant st old le emperatur s. gy f[#ur erTed

fErupto urVprovided (1) no operations /-84"N "The RHR pump may DeV --- :

are permitted that would cause dilution of the Reactor Coolant System boren

/-/ 6-A concentration, and (2) core outlet temper ure is maintained at least 10"F below saturation temperature.

/-o T-L.S

- rnsen g g W ,, gc.fc,// '!~ ~M CALLAWAY - UNIT 1 3/4 4-5 / / embreIb74/8[

4 CHANGE NUMBER H21C DESCRIPTION 1-12 H The Actions are changed to separate the required actions for only one required RHR loop OPERABLE and no required RHR loops OPERABLE. These revised Actions are consistent with the Actions which are required under this LC0 in NUREG 1431 Rev. 1, and are more conservative than current i required Actions. i 1 13 M An additional restriction is placed on the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months allowance to remove all RHR pumps from operation. This restriction precludes further draining of the RCS which is an acceptable, conservative action. This change is in l conformance with NUREG 1431 Rev. 1.

1 14 LS 22 and Action b are revised to requi OPERABLE w1 s in opera n the rod control '

system is capable of ro al and one loop in operation whe od control syst a able of rod wal. This change is consistent with 3

v. 1. rNJStr Sh-3A (3.+S-/

1 15 H A steam generator (SG) level corresponding to 10% of the wide range does not cover all of the SG tubes. To qualify as a valid heat sink, the tubes must be covered. This is a more restrictive change. [Four percent of the narrow range span is specified at the higher temperatures of MODES 3 and 4 whereas-6M of the wide range span is $ S.f,6-2 specified for MODE 5. th values ensure SG tubes are .

covered. The E0Ps cite he 4% narrow range level to ensure heat sink adequ y.]

44%

1 16 A Consistent with the intent of traveler TSTF 153, this change revises the note that permits up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months "deenergization" of RCP/RHR pumps. The revised wording clarifies the intent of the note to allow the pumps to be

" removed from operation" instead of "deenergized", thus permitting other means of removing the pumps from service.

With this change the pumps are not recuired to be deenergized to use the note (e.g., the pumps may be isolated, etc.). The change is considered to be administrative because from the standpoint of providing en exception to the LC0 requirements (to maintain the operability and operation of the pumps), the revised wording is equivalent.

1-17 - Not applicable to Callaway. See Conversion Comparison

'. Table (Enclosure 3B).

1 18 Not used.

DESCRIPTICN OF CHANGES TO CURRENT TS 3 5/15/97

RCS Loops -HODE 5. Loops Filled 3.4.7 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.7 RCS Loops-HODE 5. Loops Filled LCO 3.4.7 One residual heat removal (RHR) loop shall be OPERABLE and in operation, and either:

a. One additional RHR loop shall be OPERABLE: or

b. The secondary side ".o an water level of at least 3 steam :: B :-

generators (SGs) shall be a M 3,4 44

................................... P...............R.'t.t.?.........

1. The RHR pump of the loop in operation may be i crarsk;d E3.4 01:

1n e .: for s 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided:

a. No operations are permitted that would cause reduction of the RCS boron concentration: and

b. Core outlet temperature is maintained at least 10*F below saturation temperature.

2. One required RHR loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided that the other Rm loop is OPERABLE and in operation.

3. No reactor coolant ptmp shall be started with one-ee-mere 3 RCS EDJ cold leg temperatures s E752F M unless the secondary side water " B PS --

temperature of each SG is s g'F above each of the RCS cold leg ~Bo temperatures.

4. All RHR loops may be removed from operation during planned heatup to HOCE 4 when at least one RCS loop is in operation.

APPLICABILITY: H0DE 5 with RCS loops filled.

HARK UP OF WOG STS REV 1 (NUREG.1431) 3.4 12 5/15/97

RCS Loops -H00E 5. Loops Filled 3.4.7 ACTIONS CONDITION REWIRED ACTION COMPLETION TIME l

A. One R m loop inoperable. A.1 Initiate action to restore Immediately I a second RR loop to l AE1 OPERABLE status. I Required SGs secondary 2 side water levels rot within limits. A.2 Initiate action to restore Immediately required SG secondary side water levels to within ,

limits. '

e B. Required RHR loops B.1 Suspend all operations Immediately inoperable. involving a reduction of RCS boron concentration.

2 l

No R M loop in operation. B.2 Initiate action to restore Ismediately one RE loop to OPERABLE status and operation. l 1

SURVEILLANCE REQUIREMENTS SURVEILLANCE FRE@ENCY SR 3.4.7.1 Verify one Rm loop is in operation. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.4.7.2 Verify SG secondary side s. - o . water level 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 3.4 44 is :n We in required SGs.

U% 834.62 SR 3.4.7.3 Verify correct breaker alignment and indicated 7 days power are available to the required RHR pump that is not in operation.

MARK UP OF WOG STS REV 1 (NUREG 1431) 3.4 13 5/15/97

l RCS Loops - H00E 3 B 3.4.5 J

BASES l

ACTIONS D.1. D.2. and 0.3 (continued)

If twe RCS loops are inoperable or no RCS lop is in operation, except as during conditions permitted by the Note in ]

section, )

'i ' '

I all CRDHs r ;t t,; i ;r.;rgized by opening the RTB . or de energizing HG set $ g,%<n4 f,,e,y,.,,

All operations involving a reduction of RCS boron concentration <

must be suspended, and action to restore one of the RCS loops toggg/

OPERABLE status and operation must be initiated. Boron dilution requires forced circulation for proper mixing, and

5;r.ir.; tt.; Os ;r i cr.;r;;;ir.; tt.e = sets l

. removes the possibility of an inadvertent rod withdrawal. The immediate Completion Time reflects the importance of maintaining operation for heat removal. The action to restore must be continued until one loop is restored to OPERABLE status and L operation.

SURVEILLANCE SR 3.4.5.1 REQUIREENTS This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that the required loops are in operation. Verification includes flow rate, temperature, and 3 pump status monitoring, which help ensure that forced flow is providing heat removal. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient considering other indications and alarms available to the operator in the control room to monitor RCS loop performance.

SR 3.4.5.2 SR 3.4.5.2 requires verification of SG OPERABILITY. SG OPERABILITY is verified by ensuring that the secondary side narrow range water level is a W ~ - - - for required RCS loops. If the SG secondary side narrow range water level is 6 7.f.92

< W W he tubes may become uncovered and the associated loop may not capable of providing the heat sink for removal of the decay hea . The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of other ndications available in the control room to alert the operato to a loss of SG level.

DWEtr 0 3'.+q (continued)

MARK.UP OF NUREG 1431 BASES B 3.4 26 5/15/97

?

INSERT B 3.4-26 Q 3.4.5-2

.a 4; - t . ,

'+[$:1' l- tThe wide range level instrumentation may be used in MODE 3 in the event all narrow range

(. level instrumentation were unavailable for a required steam generator.

I i

i L

+

1 l

l I

! I L

i i

i a

- -+ * -- - - - - -

m- e.--m -

r - -

RCS Loops HODE 4 B 3.4.6 BASES (continued)

SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that one RCS or RHR loop is in operation. Verification B includet flow rate, I temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient considering other indications and alarms available to the operator in the control room to monitor RCS and RHR loop performance.

SR 3.4.6.2 SR 3.4.6.2 requires verification of SG OPERABILITY. SG OPERABILITY is verified by ensuring that the secondary side narrow range water level is := a m If the SG secondary side narrow range water level is

<m- fhetubesmaybecomeuncoveredandtheassociatedloop may rot be capable of providing the heat sink necessary for removal o decay heat. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate n view of other indications available in the control room to < 1ert the operator to the loss of SG 1evel.

A .TNf4DtT b' .T.4-32 0 T.W3 SR 3.4.6.3 Verification that the required pump is OPERABLE ensures that an additional RCS or RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.

Verification is performed by verifying proper breaker alignment and power available to the required pump. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.

REFERENCES Noner 3 m29;M l

MARK UP OF NUREG 1431 BASES B 3.4 32 .. 5/15/97

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

l.

~

[ ,

INSERT B 3.4 Q 3.4.5-2 I .4 i

L'

~

I I

The wide range level instrumentation may be used in MODE 4 in the event all narrow range !

levelinstrumentation were unavailable for a required steam generator. !

I J

l.

1 I

i l

l j

4 I

l 1

1 J

l i

l l

l.

I

__ , _ . ._ . . ~ . _ _ . _ _ _ _ _ __ _ _ _ _ _ .. . _ _ ._ ___.._ _ _ _ _ _

i. l RCS Loops H0DE 5, Loops Filled

) B 3.4.7

<v .

B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.7 RCS Loops H00E 5. Loops Filled i i j BASES l

BACKGROUND In H00E 5 with the RCS loops filled, the primary function of the l reactor coolant is the removal of decay heat and transfer 3 this heat either to the steam generator (SG) secondary side coolant

- + - .- o or the component cooling water j via the residual heat removal (RM) heat exchangers. While the j principal means for decay heat removal is via the RHR System, the J

SGs .-- .. >

- are specified as a backup means for i redundancy. Even though the SGs cannot produce steam in this 4

MODE, they are capable of being a heat sink due to their large b contained volume of secondary water. As long as the SG secondary

) side water is at a lower temperature than the reactor coolant,

! heat transfer will occur. The rate of heat transfer is directly proportional to the temperature difference. The secondary :

i function of the reactor coolant is to act as a carrier for

] soluble neutron poison, boric acid.

In MODE 5 with RCS loops filled, the reactor coolant is circulated by means of two RM loops connected to the RCS, each loop containing an Rm heat exchanger, an RHR pump, and i

appropriate flow and temperature instrumentation for control, protection, and indication. One Rm pump circulates the water i through the RCS at a sufficient rate to prevent boric acid

{ stratification . :- " - . - . -

m i

j The number of loops in operation can vary to suit the operational l

needs. The intent of this LCO is to provide forced flow from at least one Rm loop for decay heat removal and transport. The flow provided by one RHR loop is adequate for decay heat removal.

The other intent of this LCO is to require that a second path be available to provide redundancy for heat removal.

The LCO provides for redundant paths of decay heat removal capability. The first path can be an RHR loop that must be OPERABLE and in operation. The second path can be another OPERABLE RHR loop or maintaining two SGs with secondary side gg

- w water levels above H# o provide an alternate method for decay heat removal 4 /,,% O 74:f-. 2.

(continued)

MARK UP OF NUREG 1431 BASES B 3.4 33 5/15/97

RCS Loops H0DE 5 Loops Filled B 3.4.7 w ,,

BASES (continued)

APPLICABLE In MODE 5 RCS circulation is considered in the determination of SAFETY ANALYSES the time available for mitigation of the accidental boron

dilution event. ...,. ..., . - r. r. . .. - .... ... m .. ...... j

....- .. ,.,.c. ..- .

i l

i j RCS Loops H0DE 5 (Loops Filled) Mv; k;n ider.tifi;d in th; lt',0

. . . . . . , . , . . ... . . y . ..... .....m . . .om-

... 7 m a

r- C&% Q 3.+.5;-2 LCO The purpose of this LCO is to require that at least one of the Rm loops te OPERABLE and in operation with an additional RFR j loop OPERAE or two SGs with secondary side water

~

level 2 &# One Rm loop provides sufficient forced circu ation to perform W safety functions of the reactor coolant under these conditions An additional Rm loop is required to be OPERABLE to : single failure considerations.

However, if the standby RIR 1 p is not OPERABLE, an acceptable alternate method is two SGs th their secondary side m water levels a W $. Sho d the operating RHR loop fail, the SGs could be us to remove he decay heat -

M-~v

~

my y,_g g 747 3 Note 1 permits all Rm pumps to be i crergic;d M j M s I hour per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. The purpose of the Note is j to permit tests a o o . .'

u . . .

-?. ~~ .... m

) m i;isc.cd t; veliit; v;rieus eccii..; ear.ly;u v;le; . CT,; cf tk t;;t gifernd 17176 tM ;tertup t;;tirs pregic. is the veliitica of red diep tins durir,; ;;1d j ;;aditiens. kth with er,d witkut now. O.; a; new test ;y k j ps.crerad in it0^C 3. 4. ;r and requir;s that the pu ps M

(continued)

I

HARX UP OF NUREG 1431 BASES B 3.4 34 5/15/97 i

l

INSERT B 3.4-34 Q 3.4.5-2 3:;+.

'- The narrow range level instrumentatian may be used in MODE 5 if the one wide range level indicator per steam generator were unavailable.

. (,

d o

I I

re I

RCS Loops H0DE 5, Loops Filled B 3.4.7 BASES LCO RHR pumps are OPERABLE if they are capable of being powered and (continued) are able to provide flow if required. An OPERABLE SG can perform as a heat sink nummunnmemmmmmmmmmun when it has an adequate water level and is OPERABLE in accordance with the Steam Generator Tube Surveillance Program.

APPLICABILITY In H00E 5 with RCS loops filled, this LC0 requires forced circulation of the reactor coolant to remove decay heat from the core and to provide proper boron mixing. One loop of RHR provides sufficient circulation for these purposes. However, one additional RHR loop is required to be OPERABLE, or the secondary side water level of at least SGs is required to be 2 ig Q S.4,5%2 Ope tfoninotherMODESiscoveredby:

LCO 3.4.4, "RCS Loops H0 DES 1 and 2":

LEO 3.4.5, "RCS Loops H0DE 3":

LCO 3.4.6, "RCS Loops H0DE 4";

LCO 3.4.8, "RCS Loops H00E 5 Loops Not Filled";

LCO 3.9.5, " Residual Heat Removal (RM) and Coolant Circulation High Water Level" (H00E 6); and LCO 3.9.6, " Residual Heit Removal (RHR) and Coolant Circulation Low Water Level" (HODE 6).

ACTION? A.1 and A.2 g

If one RIE loop is inoperable and the required SGs have secondary side m water levels < #$ redundancy for heat removal is lost. Action must be initiated immediately to restore a second RHR loop to OPERABLE status or to restore the required SG secondary side water levels. Either Required Action A.1 or Required Action A.2 will restore redundant heat removal paths.

The immediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal.

B.1 and B.2 If no RHR l'oop is in operation, except during conditions permitted by Note [ 1 !.w , or if no loop is OPERABLE, all (continued)

MARK UP OF NUREG 1431 BASES B 3.4 36 5/15/97 l

RCS Loops H00E 5. Loops Filled B 3.4.7 i

i BASES ACTIONS B.1 and B.2 (continued)

]

operations involving a reduction of RCS boron concentration must be suspended and action to restore one RHR loop to OPERABLE

]i status and operation mu:;t be initiated. To prevent 6

- boron dilution, forced circulation M 3 -

is required to provide proper mixing. ;rd i;r;.;;. s; i

tt.; ergin t; critic;11ty in tt.i; tg; cf ; rcr;ti;n. The immediate Completion Times reflect the importance of maintaining i operation for heat removal.

1 i

SURVEILLANCE SR 3.4.7.1

) REQUIREENTS This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that the required

loop it in operation. Verification includet flow rate,

' temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal. The Frequency of,12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient considering other indications and alarms available 1

to the operator in the control room to monitor RHR loop performance.

}

{ SR 3.4.7.2 g (,(,0f i

i Verifying that at least two SGs are OPERABLE by ensurirptheir i secondary side nettow M range water levels are m &#M $ 3M-2.

ensures an alternate decay heat removal method E

)

1 in the event that the second Rm loop is not OPERABLE. If both RHR loops are OPERABLE, this l

Surveillance is not . The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of othe indications available in the control room to alert the oper r to the loss of SG level.

.rhiSGC7" $ 3$-27 & 2 f.90-SR 3.4.7.3 Verification it.'t a second RM pump is OPERABLE ensures that an additional pug can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.

Verification is performed by verifying proper breaker alignment and power available to the Rm pump. If secondary side . <

g water level is a W g in at least two SGs, this f)3.fS-2 4(.%

(continued)

MARK UP OF NUREG 1431 BASES B 3.4 37 5/15/97

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

l INSERT B 3.4-37 Q 3.4.5-2 L.

[- The narrow range levelinstrumentation may be used in MODE 5 if the one wide range level indicator per steam generator were unavailable.

f.

i.

t I

T,

+ .

l' l

l l'

.--- , y - - - -

+-v ,e ' -r

CHANGE )

NUPEER JUSTIFICATION 3.4 42 Not applicable to Callaway. See Conversion Comparison Table (Enclosure 68).

3.4 43 Not applicable to Callaway. See Conversion Comparison Table (Enclosure 68).

Steam generator lev s for MODES 3. 4, and 5 are specified to ensure SG Oons, w tubes are covered. The ";11:x; current TS did ensure tube f) f, f. fP-2 f *I"Y ? 2 , overage. Afo}e uder

~

3.4 45 BITS 3.4.12 has been revised to move the No:e for Required Action B.1 regarding CCP pump swap operatio and theVApplicability "-te f:r g S, f, /2--2 reg *./T7" accumulator isolation to the LCO ,M dir-M " tr=:ict "^^ SI q f y ,n

.x 1.glant specific time allowances for exceeding the LCO's number 7h3' N# Mima of [ECCS] pumps capable of injecting into the RCS are incorporated [, as lean rem,& "ater detti' ri+"eti::: ut . . l

*/ [ discussed in CN 3.4 18].

=nf.irre tr'tb LC^ r: p:ritt:d oud hm au r; :;;- ;-i:t ly k ey, g ' x xt:t:d r i r tb LCO.

LCo,4*for*vh t

A/.r,, J.4 46 Consistent with current TS 3/4.1.1.4, " Minimum Temperature for Criticality " ITS LCO 3.4.2 and its Condition A and SR 3.4.2.1 are modified to refer to " operating" RCS loops. Adopting the current TS wording is acceptable since valid T m measurements are not obtainable for a non operating loop.

3.4 47 ISTS SR 3.4.11.1 contains a Note which exempts the cycling of the block valve when it is closed in accordance with Required Actions of Conditions B or E of LCO 3.4.11. However. Required Action A.1 also directs closure of the block valve when one or more PORVs are inoperable and capable of being manually cycled. The SR Note should also exempt performance when the block valve is closed in accordance with Required Action A.1 as the block valve should not be opened when the PORV is inoperable. This change is consistent with NUREG 1430 and NUREG 1432 inasmuch as the block valve cycling is exempted under Conditions A, B, and E. --Si=: p r- t: tb bl x k ^;;1::':} i: $ 3',f.//-+

= int:id ir. ".:;;ir;d Acties A.1,75e Note to SR 3.4.11.1 will be 4 g ,7y,( revised to not require the surveillance performance if the block Achs g vags{_}sggs pefg}:]} g.jg _ygy{ggk_ g)) _- g g, e

                                .. . _ ..,. ... . w n - ~......o.m         m.  . . .   . ~ . . . . . . . . ~ . . . . . . - . ~

_ _ eh- t% rdi~; d=;; 'rs "at" t: "perfernd" 4- tt Ste. It,o ;rdin; Of S" 3.4.11.1 ie redaed te acc- ;d;t; tt.; C=diti= S Sad E a---ti= This change is consistent with traveler WOG 87. ;

                                                   $ rNNA1~ M-7A 3.4 48          A note is added to ITS 3.4.8 ACTIONS indicating that entry into MODE 5                                  )

Loops Not Fille.d from MODE 5 Loops Filled is not permitted while ! LC0 3.4.8 is not met. The addition of this note is based on the j performance of a plant specific LCO 3.0.4 matrix (see CN 1-02 LS-1 of the CTS 3/4.0 package). ;ZA/Mg7- 6A-78 4 1',f A / l l JUSTIFICATION FOR DIFFERENCES TS 7 5/15/97

l T 4

                   ' r yer   iah,rr o.ajhaj.p._ yr s ygf3 y DIFFERENCE FROM NUREG-1431                                                                                                                                                          APPLICABILITY NUPEEL    DESCRIPTION                                                                                                                                                DIABLO CANYON        COMANCHE PEAK           WOLF CREEK                                                                                      CALLAWAY 3.4-37    The primary to secondary leakage limits are revised per                                                                                                    No                   No                      No                                                                                              Yes Callaway OL Amendment No.116 dated October 1,1996.

3.4-38 emme r t--t ,,;;;, a m y n;r ;;;, um g;;n ; ;g go -yes- -yes-- _yes.

           ^
                    . .         L ;     .

N h $*NN 3.4-39 The shutdown requirements of ITS 3.4.11 would require the Yes Yes Yes Yes plant to reduce T to < 500*F witt.in 12 hours, rather than go to MODE 4. to address the concern of entering [COMS]

LCO 3.4.12 Applicability with inoperable PORVs. For consistency, the shutdown requirements of ITS 3.4.16 are also revised to allow 12 hours to reduce T, to < 500*F. This change is consistent with traveler TSTF-113. 3.4-40 The Note to SR 3.4.1.4 would be modified to specify a No - See CN 3.4-51. No - See CN 3.4-34. Yes Yes plant specific reactor power and to provide additional time to perform an RCS precision flow rate measurement. 3.4-41 LCO 3.4.1 is revised to reference Tables 3.4.1-1 and Yes - Allowance No No No 3.4.12 for RCS total flow rate limits for DCPP Units 1 added per Amendment and 2 respectively. 60/59. i 3.4 42 An exception to SR 3.4.14.1 frequency to leak test Yes No No No PIVs 8802A 8802B and 8703 has been added. This change is consistent with the DCPP current TS. 3.4-43 A new Condition C is added to LCO 3.4.1 to reflect the No No Yes No current TS of Wolf Creek for RCS flow rate. 3.4 44 Steam generator levels for MODES 3. 4. and 5 are specified No No

                                                                                                                                                                                                                 -Sie-[e.r                                                                                        Yes g g g _7 to ensure SG tubes are covered. The Ol? x ; current TS did not ensure tube coverage.

CONVERSION COMPARISON TABLE NUREG-1431 5/15/97

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.5-3 APPLICABILITY: CA, WC l REQUEST: CTS 4.4.1.2.2,4.4.1.3.2 and 3.4.1.4.1.b and ITS 3.4.5,6 and 7 (Callawey and Wolf Creek) Comment: Ten percent wide range level was specified as the necessary heat sink level. Now in the ITS the levelis narrow range. Was this is a known error in the TS that is nov' being corrected or was this just discovered as part of the conversion effort? Please piUvide the technical basis foi concluding that 10% (4% for Callaway) narrow range is adequate. Additionally explain why different narrow range level values are used at each plant and why wide range level is used in Mode 5 at one and not the other. FLOG RESPONSE: The CTS Bases for Reactor Coolant Loops and Coolant Circulation is silent on the background behind the 10% wide range value specified. During the ITS conversion process, industry traveler TSTF-114 was developed (and subsequently approved by NRC), to recognize the importance of keeping the SG tubes covered as discussed in NRC IN 95-35, " Degraded Ability of SGs to Remove Decay Heat by Natural Circulation." Since 10% on the wide range level instrumentation does not ensure the SG tubes are fully covered, the values were changed in the conversion amendment. This was not an error known prior to starting the amendment development for ITS. Plant Specific Discussion The basis for using 4% narrow range for Callaway in MODES 3 and 4, as well as using 67% wide range for MODE 5, is discussed in response to Comment Number 3.4.5-2. The 4% narrow range value was chosen since it is used throughout the EOPs, it has operator awareness because of the EOP familiarity, and ensures an SG water level approximately 100 inches above the top of the highest SG tube. Wide range levelinstrumentation was chosen for MODE 5 since it is calibrated for cold conditions and provides a larger span. ATTACHED PAGES: None

1 l i ADDITIONAL INFORMATION COVER SHEET l ADDITIONAL INFORMATION NO: Q 3.4.6-1 APPLICABILITY: CA, CP, DC, WC REQUEST: Difference 3.4-02 Comment: The difference states that the STS doesn't cover all possible configurations I and the language of the STS is potentially confusing. Pleasa explain the basis for these comments. FLOG RESPONSE: The STS wording for Condition A, "One required RCS loop inoperable AND Two RHR loops inoperable", and for Condition B, "One required P.HR loop inoperable AND Two required RCS loops inoperable", is confusing. This confusion arises from the fact the l LCO allows any combination of two RCS or RHR loops, including one RCS loop and one RHR l loop, to satisfy the OPERABILITY requirement yet Conditions A and B are worded as if either two RCS loops or two RHR loops, exclusively, were the required loops. By way of illustration, the following scenarios are presented. Assume the LCO's OPERABILITY requirements are satisfied by one RCS loop and one RHR loop. These loops are serving as the " required" loops. If the RCS loop becomes inoperable, Condition A does not apply because j it is "ANDED" with "Two RHR loops inoperable" yet one RHR loop remains OPERABLE in this scenario. Conversely, if the RHR loop becomes inoperable, Condition B does not apply because it is "ANDED" with "Two required RCS loops inoperable" yet one RCS loop remains OPERABLE. In fact, the wording of STS Condition B is at odds with the LCO since Condition B requires three loops to be OPERABLE (one RHR and two RCS loops). The FLOG considered this wording to be a potential source of error for plant operators. Since the corresponding CTS specification is not confusing it was adopted in lieu of the STS wording. This confusion also led to the WOG creating a traveler, WOG-109, which was subsequently withdrawn and superseded by TSTF-263 which is currently under NRC review. TSTF-263 presents a very similar approach to that used by the FLOG to correct STS 3.4.6; however, TSTF-263 has not been incorporated by the FLOG. TSTF-263 was not issued until several months after the FLOG submittals. The changes incorporated in ITS 3.4.6 are based I on the CTS which has less rigid logic connectors than the STS. ATTACHED PAGES: None i I l

__ . _ - . __= _ l ADDITIONAL INFORMATION COVER SHEET l ADDITIONAL INFORMATION NO: Q 3.4.7-3 APPLICABILITY: CA l REQUEST: ITS Bases 3.4.7 Background (Callaway) Comment: The last paragraph on smooth Bases Page B 3.4-32 incorrectly states ". l above 7E" This error does not appear in the highlight / strikeout version of the Bases. i

FLOG RESPONSE: This error is corrected in the attached page. This error had already been identified during initial operator training using the " smooth"ITS. Additional typing errors ) and editorial corrections are included in Additional Information Number CA-3.4-002. ATTACHED PAGES: l l Attachment 20, Smooth Bases Page B 3.4-32

m._ >_. - _ _ . _ . _ . _ . _ . _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ . _ _ _ . _ . _ _ _ _ _ _ _ , RCS Loops MODE 5. Loops Filled j B 3.4.7 B 3.4 REACTOR COOLANT SYSTEM (RCS) I B 3.4.7 RCS Loops MODE 5, Loops Filled BASES BACKGROUND In MODE 5 with the RCS loops filled, the primary function of the reactor coolant is the removal of decay heat and transfer of this heat either to the steam generator (SG) secondary side coolant via natural circulation (Ref.1) or the component cooling water via the residual heat removal (RIR) heat exchangers. While the principal means for decay heat removal is via the RlR System, the SGs via natural circulation are specified as a backup means for redundancy. Even though the SGs cannot produce steam in this MODE, they are capable of being a heat sink due to their large contained volume of secondary water. As long as the SG secondary I side water is at a lower temperature than the reactor coolant. I heat transfer will occur. The rate of heat transfer is directly proportional to the temperature difference. The secondary i function of the reactor coolant is to act as a carrier for soluble neutron poison, boric acid. , l In MODE 5 with RCS loops filled, the reactor coolant is circulated by means of two RIR loops connected to the RCS, each loop containing an RlR heat exchanger, an Rm pump, and appropriate flow and temperature instrumentation for control, protection, and indication. One Rm pump circulates the water through the RCS at a sufficient rate to prevent boric acid stratification but is not sufficient for the boron dilution analysis discussed below. The number of loops in operation can vary to suit the operational needs. The intent of this LCO is to provide forced flow from at least one RHR loop for decay heat removal and transport. The flow provided by one RHR loop is adequate for decay heat removal. The other intent of this LCO is to require that a second path be

                                                      ' available to provide redundancy for heat removal.

The LC0 provides for redundant paths of decay heat removal capability. T.m first path can be an RlR loop that must be OPERABLE and in operation. The second path can be another OPERABLE RHR loop or maint ining two SGs with secondary side wide range water levels above to provide an alternate method for CA U C decay heat removal via na ural circulation. ' @ J,4 7'S

                                                                                                      & b$0/o 03M.5-3 (continued)

CALLAWAY PLANT ITS BASES B 3.4 32 5/15/97

____._...__m . _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ . _ . . _ . . _ . _ . _ _ j ADDITIONAL INFORMATION COVER SHEET l ADDITIONAL INFORMATION NO: Q 3.4.8-1 APPLICABILITY: CA, CP, DC, WC REQUEST: Difference 3.4-48 Comment: it is unclear why TS 3.0.4 would not apply. If this change is to be considered it should be done on a generic basis. I I !' FLOG RESPONSE: A Reviewer's Note in STS LCO 3.0.4 states: "LCO 3.0.4. has been revised l so that changes in MODES or other specified conditions in the Applicability that are part of a

               ' shutdown of the unit shall not be prevented. In addition, LCO 3.0.4 has been revised so that it

. is only applicable for entry into a MODE or other specified conditions in the Applicability in ! MODES 1,2,3, and 4. The MODE change restrictions in LCO 3.0.4 were previously applicable in all MODES. Before this version of LCO 3.0.4 can be implemented on a plant-specific basis,

the licensee must review the existing technical specifications to determine where specific restrictions on MODE changes or Required Actions should be included in individual LCOs to justify this change; such an evaluation should be summarized in a matrix of all existing LCOs

                'to facilitate NRC staff review of a conversion to the STS " Based on this Reviewer's Note, a j                  matrix of this evaluation was placed in the NSHC LS-1 in Enclosure 4 of the Section 3.0 i                  package (Attachment No. 6).

JFD 3.4-48 has been revised to incorporate additional justification from NSHC LS-1 in Enclosure 4 of the Section 3.0 package (Attachment No. 6). JFD 3.4-48 has l been revised to include: l l l "LCO 3.0.4 has been revised so that changes in MODES or other specified conditions in the l Applicability that are part of a shutdown of the unit shall not be prevented. In addition, LCO j 3.0.4 has been revised so that it is only applicable for entry into a MODE or other specified

conditions in the Applicability in MODES 1, 2, 3, and 4. The MODE change restrictions in LCO 3.0.4 were previously applicable in all MODES. ITS LCO 3.4.8 is modified by a Note stating: l l

                 "While this LCO is not met, entry into MODE 5, Loops Not Filled from MODE 5, Loops Filled is I

not permitted." The transition from MODE 5 (loops filled) to MODE 5 (loops not filled) removes the steam generators as a decay heat removal system while the RHR System is potentially

                 . degraded. Therefore, the Note ensures that the transition is precluded if LCO 3.4.7.b (two l                  SGs) were chosen (in lieu of the second RHR loop) to ensure decay heat removal capability

! ' prior to draining +he RCS."

               - It should be noted that the Applicability Bases for ITS 3.4.8 already provides a similar h                 discussion.

ATTACHED PAGES: 4 f Attachment 10, CTS 3/4.4 - ITS 3.4 l Enclosure 6A, page 7 a 1

CHANGE MutBER JUSTIFICATION 3 .4 4:' Not applicable to Callaway. See Conversion Comparison Table i (Enclosure 6B). ! 3.4 43 Not applicable to Callaway. See Conversion Comparison Table (Enclosure 68).

                       .4 44              Steam generator levels for H00ES 3, 4, and 5 are specified to ensure SG bn                    "'

tubes are covered. The C;1 h n; current TS did not ensure tube f) 7. f. f-2 1

        & *I'Y M 2 ,, overage.                                                                   M.{e uder                                   j 3.4 45      blTS 3.4.12 has been revised to move the No e for Required Action B.1 regarding CCP pump swap operatio                and t Applicability St f:r           g 7,f;/2-l:t reg MT7"raccumulator isolation to the LC0 ,M Picurd " t=:lcr E 51- O 7.y ;g 15./lant specific time allowances for exceeding the LCO's number
       %38 Nehe$m                         of [ECCS] pumps capable of injecting into the RCS are incorporated [. as Id*" NMd2 0'                       discussed in CN 3.4 18]. A re umtet get:$, gitgete = ,,3;7;

=:;tirnr te tb LC^ :r: pe-itt;d and m c er; ;; eprht:ly ene fe 1M

n=t:t:d ud:r th: LCO.

LCo t< /*r*v'&s_j3 A/f,, 3.4 46 Consistent with current TS 3/4.1.1.4, "Hinimum Temperature for Criticality," ITS LCO 3.4.2 and its Condition A and SR 3.4.2.1 are modified to refer to " operating" RCS loops. Adopting the current TS wording is acceptable since valid T,, measurements are not obtainable for a non operating loop. 3.4 47 ISTS SR 3.4.11.1 contains a Note which exempts the cycling of the block valve when it is closed in accordance with Required Actions of Conditions B or E of LC0 3.4.11. However, Required Action A.1 also directs closure of the block valve when one or more PORVs are inoperable and capable of being manually cycled. The SR Note should also exempt performance when the block valve is closed in accordance with Required Action A.1 as the block valve should not be opened when the PORV is inoperable. This change is consistent with NUREG 1430 and NUREG 1432 inassoch as the block valve cycling is exempted under Conditions A, B, and E. --Si=0 m: r to the bbck nh C:) h S 7,f,//- + nint:ined in ".:quir:d Actien A.1,'ETie Note to SR 3.4.11.1 will be 4 g ,7aj revised to not require the surveillance performance if the block p g valve (s) is closed pef'C:nditi = ^ Six; penr t; the blxk =h (:) hg u r;.x xd in rewiced Actien; ".2 m,d E.0, the arxilh=: =n =t bc-

                                           =+      cha +% erding cbn; fre; "at" t: "p^rf =d"                    " the "Otc.-

thu .;;rdt; cf S" 3.4.11.1 is i evised t; n;a ed;t: the C;nditi n S This change is consistent with traveler WOG 87. H E emti:n b rNMA'r M-7A 3.4 48 A note is added to ITS 3.4.8 ACTIONS indicating that entry into MODE 5 Loops Not Filled from H00E 5 Loops Filled is not permitted while LCO 3.4.8 is not met. The addition of this note is based on the performance of a plant specific LCO 3.0.4 matrix (see CN 102 LS-1 of the CTS 3/4.0 package). %A/KF 6A-78 g M /-/ JUSTIFICATION FOR DIFFERENCES - TS 7 5/15/97

INSERT 6A-78 Q 3.4.8-1 LCO 3.0.4 has been revised so that changes in MODES or other specified conditions in the Applicability that are part of a shutdown of the unit shall not be prevented. In addition, LCO 3.0.4 has been revised so that it is only applicable for entry into a MODE or other specified conditions in the Applicability in MODES 1,2,3, and 4. The MODE change restrictions in LCO 3.0.4 were previously applicable in all MODES. ITS LCO 3.4.8 is modified by a Note stating: "While this LCO is not met, entry into MODE 5, Loops Not Filled from MODE 5, Loops Filled is not permitted." The transition from MODE 5 (loops filled) to MODE 5 (loops not filled) removes the steam generators as a decay heat removal system while the RHR System is potentially degraded. Therefore, the Note ensures that the transition is precluded if LCO 3.4.7.b (two SGs) were chosen (in lieu of the second RHR loop) to ensure decay heat removal capability prior to draining the RCS. i

,n

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.4.9-1 APPLICABILITY: CA, CP, DC, WC REQUEST: ITS 3.4.9 Comment: Does 92% (90% for Diablo Canyon) in the pressurizer ensure that upon an inadvertent Si that the pressurizer will not overfill before the operator is assumed to take action? Other plants have lowered this limit (Robinson) or qualified the PORVs for water (Millstone 3). FLOG RESPONSE: ITS Surveillance Requirement 3.4.9.1 requires the pressurizer water level to be less than 92% (90% for Diablo Canyon). This requirement is not related to the assumptions used in the inadvertent safety injection analysis. The basis for this requirement is given in the ITS Bases for SR 3.4.9.1 (as clarified by NRC approved TSTF-162), which states that it is to ensure provision of a minimum space for a steam bubble which is an assumption in the safety analyses (i.e., the pressurizer must not be water solid). This maximum pressurizer levelis not assumed in any safety analysis. ATTACHED PAGES: None

r l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.4.9-2 APPLICABILITY: CA l l REQUEST: ITS LCO 3.4.9.b (Callaway) Comment: The ITS should read " .150 Kw." l FLOG RESPONSE: This error is corrected in the attached page. Similar to Comment Number 3.4.7-3, this error had already been identified during initial operator training using the " smooth" ITS. This error does not appear in the highlight / strikeout version of the ITS. Additional typing errors and editorial corrections are included in Additional information Number CA-3.4-002. ATTACHED PAGES: Attachment 19, Smooth ITS Page 3.4-16 l l l l

Pressuriztr 3.4.9

  .3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 Pressurizer LCO 3.4.9             The pressurizer shall be OPERABLE with:

a. Pressurizer water level s 92*: and

b. Two groups of backup pressurizer heaters OPERABLE with the capacity of each group 2150 (. ' A r. ? v,i W d 8 f.9-2.

APPLICABILITY: MODES 1, 2 and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME i A. Pressurizer water level A.1 Be in MODE 3. 6 hours not within limit. M A.2 Fully insert all rods. 6 hours M A.3 Place Rod Control System in 6 hours l a condition incapable of rod l withdrawal. { M l A.4 Be in MODE 4. 12 hours 1

l. B. One required group of B.1 Restore required group of 72 hours

! backup pressurizer backup pressurizer heaters heaters inoperable. to OPERABLE status. (continued) l CALLAWAY PLANT ITS 3.4 16 5/15/97

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

1 ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.4.10-1 APPLICABILITY: CA, CP, DC, WC l REQUEST: ITS 3.4.10 Bases Applicable Safety Analyses Comment: What justifies the differences between the ITS Bases and the STS Bases and between the plant Bases (especially Callaway and Wolf Creek) of the lists of possible over pressurization events? FLOG RESPONSE: These Bases changes reflect each plant's licensing basis as expressed in their respective versions of FSAR Chapter 15. l 1 Plant Soecific Discussion l l In the Callaway Applicable Safety Analysis Bases for ITS 3.4.10, changes were made to items a, c, and e and new items g and h were added. The change to item a is strictly editorial, consistent with the discussion of this event in FSAR Section 15.4.2. The change to item c is consistent with the discussion in FSAR Sections 15.2.2 and 15.2.3, i.e., no loss of extemal I load analysis is presented in FSAR Section 15.2 since the turbine trip event is more limiting. ' The change to item e denotes that this event is not a station blackout; standby power is available from the diesel generators consistent with the discussion in FSAR Section 15.2.6. Feedwater line break is added as new item g since the analysis of this event shows the primary and secondary side safeties lift, as discussed in FSAR Section 15.2.8 (see FSAR Figures 15.2-16 and 15.2-21 for the pressurizer pressure and volume transients). RCCA ejection is added as new item h since there is pressure surge analysis, discussed on page 15.4-41 of Section 15.4.8.2.2 that is incorporated by reference to WCAP-7588, Rev.1-A, January 1975,"An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetic Methods," Risher, D. H., Jr. This pressure surge analysis concludes that with an ejection worth of one dollar at BOL and HFP conditions, the resulting stress levels do not exceed faulted limits. Addition of this event was specifically discussed with i the Westinghouse transient analysis engineer assigned to Callaway. l All of the changes in this Bases section transform a generic discussion to one that applies to i this plant specifically. ATTACHED PAGES: None

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.4.11-1 APPLICABILITY: CA, CP, DC, WC

REQUEST: Change 4-04 LG . 1 Comment : The requirement is in the CTS and the STS. The justification for not putting i it in the ITS is that automatic actuation to open is not required. However, proper calibration also ensures that the PORV does not prematurely open creating as stated in the Bases "in effect a small break LOCA." FLOG RESPONSE: There are two CTS LCOs (3.4.9.3 (3.4.8.3 for CPSES) & 3.4.4) and corresponding ITS LCOs (3.4.12 & 3.4.11) controlling pressurizer PORV operability. One of these,' CTS 3.4.9.3 (3.4.8.3 for CPSES) and corresponding ITS 3.4.12, govems their operability as part of the LTOP/COMS system. Both the CTS (SR 4.4.9.3.1.b or 4.4.8.3.1.b for CPSES) and the iTS (SR 3.4.12.9) require CHANNEL CAllBRATIONs of the LTOP/COMS PORV actuation channels every 18 months to support this function. The second of these, CTS 3.4.4 and ITS 3.4.11 governs the operability of the PORVs and their block valves as isolable relief valves. While the ability to open the PORVs manually and to isolate a stuck open PORV using its block valve are considered safety-related capabilities, the ability of the PORVs to act as automatic relief valves in Modes 1. 2, and 3 is not a safety function in the current licensing basis. The pressurizer safeties fulfill both the RCS Code overpressure protection function and the automatic pressure relief function assumed in the accident analyses. For this reason, STS 3.4.11 does not have a CHANNEL CAllBRATION surveillance requirement. SR 4.4.4.1.b is therefore moved out of the technical specifications by DOC 4-04-LG. This is appropriate since automatic actuation of the PORVs is not a safety function in Modes 1,2, or 3. Requirements that are not needed to support the safety analyses are moved out of the Technical Specifications, reflecting the philosophy and content of NUREG-1431. l The premature opening of a PORV is considered to be a small break LOCA. A LOCA is an unisolable leak or break in the RCS. A stuck open pressurizer safety valve would constitute a LOCA. One of the design functions of the PORVs is, however, to reduce the risk of a stuck ' open safety by having actuation set points below those of the safeties. As stated in the STS LCO Bases for ITS 3.4.11, a stuck open PORV could be isolated by closing its safety related block valve, thus avoiding a LOCA. The automatic actuation of a PORV at a pressure lower than its nominal design setpoint is not desirable, but is not outside the safety analyses. ATTACHED PAGES: None 4 i ~

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.4.11-2 APPLICABILITY: CA, CP, DC, WC l REQUEST: Change 4-08 LS 34 and Difference 3.4-35 l Comment: WOG-60 has not yet become a TSTF. FLOG RESPONSE: WOG-60 has been approved by the TSTF and is designated as TSTF-288. This traveler has been submitted to the NRC and is under review. The proposed wording in TSTF-288 was modified from WOG-60, Rev.1, and these modifications have been incorporated into the ITS (editorial SR Bases Note change). The FLOG continues to pursue the changes proposed by this traveler. ATTACHED PAGES: Attachment 10, CTS 3/4.4 - ITS 3.4 Enclosure 3A, page 7 Enclosure 3B, page 5 Enclosure 4, page 58 Enclosure SA, Traveler Status Sheet Enclosure SB, page B 3.4-61 Enclosure 6A, page 6 I 1 I

CHANGE NUPBER EC DESCRIPTION steam dump] valves. This change is consistent with traveler TSTF 113, Rev.' 3. 4 06 LS 32 Not applicable to Callaway. See Conversion Comparison Table (Enclosure 38). 4 07 LS 33 Not applicable to Callaway. See Conversion Comparison Table (Enclosure 38).

                                                                                         -TTYF'-nPP 4 08                         LS-34.         Consistent with traveler 'E 50,Vthe requirement to O M, //~.2 perfonn the 92 day surveillance of the pressurizer PORV block valves and the 18 month surveillance of the pressurizer PORVS (i.e., perform one complete cycle of each valve) is revised to indicate that the surveillance is only required to be performed in H00ES 1 and 2. This is consistent with the recommendations of Generic Letter 90 06, " Resolution of Generic Issue 70, ' Power Operated Relief Valve and Block Valve Reliability,' and Generic Issue 94, ' Additional Low Temperature Overpressure Protection for Light Water Reactors,' Pursuant to 10CFR50.54(f)," June 25, 1990.

4 09 LS 36 The requirement to perform the 92 day surveillance of the pressurizer PORY block valves (i.e., perform one complete cycle of each block valve) is revised such that it is not required if the block valve is closed to meet ACTION a of current TS LC0 3.4.4. This change is acceptable because no credit is taken for the automatic actuation of the PORV in Modes 1, 2, or 3. Credit is taken for manual operation of the PORVs during the Steam Generator Tube Rupture ! (SGTR) accident. However, the capability to manually cycle the PORVs will be unaffected by this change. This change will not affect the ability of the block valve to open, if closed to meet ACTION a, in the mitigation of an SGTR. Deferral of the block valve cycling surveillance-will not diminish the design capability of the block valve to open against differential pressures that would present after an SGTR since the block valves are capable of opening against 2485 psig, the safety valve lift pressure, h , pggy __ 4ereas pressurizer pressure decreases after an SGTR. g gfg This change is consistent with traveler WOG 87. 5 01 A This change moves the steam generator tube surveillances to ITS SR 3.4.13.2 and the Administrative Controls t Sections 5.5.9 SG Tube Surveillance Program, and 5.6.10, ' SG Tube Inspection Report. DESCRIPTION OF CHANGES TO CURRENT TS 7 5/15/97

2

                                                                                                                                                                                                                                                     .a o CONVERSION COMPARISON TABLE - CURRENT TS 3/4.4                                                                                                                                                Page 5 of 13        ,

TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY 4 05 The shutdown requirements of CTS LCO 3.4.4 would require the Yes Yes Yes Yes LS-31 plant to reduce T to < 500 7 within 12 hours, rather than go to MODE 4. to address the concern of entering [COMS] LCO Applicability with inoperable PORVs. For consistency, the shutdown requirements of CTS LCO [3.4.8] would be similarly revised. 4 06 This change provides a 72 hour completion time to restore an Yes No - Already part No - Already part No - Already part LS-32 inoperable block valve, with the PORV placed in manual of current TS. of current TS. of current TS. control mode. The current TS requires the block valve to be restored within one hour, or remove power from the solenoid. 4-07 This change provides a two hour completion time for Yes No - Already part No - Already part No - Already part LS-33 restoring an inoperable block valve when more than one block of current TS. of current TS. of current TS. valve is inoperable, and 72 hours to restore the remaining - valves. The current TS requires the block valve to be restored within one hour for one or more valves inoperable. 4 08 Consistent with traveler Z 0 ; Ihe

remerbtb perform Yes Yes Yes Yes

                                                                                                                                                                                                                                                        ' UIY LS-34  the PORY and block valve cycling surveillances is revised such that the surveillances are only required to be performed in MODES I and 2.

Consistent with traveler WOG-87. the requirement to perform Yes Yes Yes Yes 4-09 LS 36 the 92 day surveillance of the pressurizer PORY block valves (i.e., perform one conplate cycle of each block valve) is revised such that it is not required if the block valve is closed to meet ACTION a of current TS LCO 3.4.4. Q a//,W ,Q ap A/ /r 34N f (cy f, grweg e,$/fy f,/s/ 7 70,2ffj_4 u n c To ornv*ernk/t IUK Ylf) P'Weo- f c~ b or- CL 5-01 This change moves the Steam Generator Tube Surveillances to #Yes No - Same as Yes 4 Yes A ITS SR 3.4.13.2 and the Administrative Controls Sections change 1-14 A for 5.5.9 and 5.6.10. CTS Saction 3/4.0. CTS LCO 3.4.5 is deleted. Steam Generator oper cility Yes No - CPSES does not Yes Yes 5-02 A requirements in MODES 1-4 are specified in the RCS lo@ snd have this leakage specifications. specification. Clarification to remove potential interpretation problems Yes No - Same as CPSES Yes Yes 5-03 A related to probe orientation versus entry point. change 1-15 for CTS Section 3/4.0. CONVERSION COMPARISON TABLE - CURRENT TS 5/15/97

IV. SPECIFIC N0 SIGNIFICANT HAZARDS CONSIDERATIONS I NSHC LS 34 10 CFR 50.92 EVALUATION

FOR TECHNICAL CHANGES THAT IMPOSE LESS RESTRICTIVE REQUIREMENTS WITHIN THE TECHNICAL SPECIFICATIONS l TD/~2Pf; 1 Consistent with travelerM0G4&the requirement to perform the 92 day surveillanced ?.4,//4 of the pressurizer PORV block valves and the 18 month surveillance of the pressurizer PORVs (i.e.,' perform one complete cycle of each valve) is revised to indicate that the surveillance is only required to be performed in MODES 1 and 2.

This is consistent with the reconmendations of Generic Letter 90 06, " Resolution of Generic Issue 70, ' Power Operated Relief Valve and Block Valve Reliability,' and l Generic Issue 94, ' Additional Low Temperature Overpressure Protection for Light- 1 Water Reactors,' Pursuant to 10CFR50.54(f)," June 25, 1990. This proposed TS change has been evaluated and it has been determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92(c) as quoted ; below:

                           "The Commission may make a final determination, pursuant to the procedures in                    !

50.91, that a proposed amendment to an operating license for a facility licensed under 50.21(b) or 50.22 or for a testing facility involves no I significant hazards consideration, if operation of the facility in accordance with the proposed amendment would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated: or

2. Create the possibility of a new or different kind of accident from any accident previously evaluated: or

3. Involve a significant reduction in a margin of safety."

The following evaluation is provided for the three categories of the significant hazards consideration standards:

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

The Pressurizer PORV LC0 requires the PORVs and block valves to be Operable for manual operation to mitigate the effects associated with a Steam Generator Tube Rupture (SGTR). The SGTR is primarily a concern during Modes 1 and 2 because the reactor is' producing heat, transients are more likely, and the 4 consequences of an accident are more severe. In Mode 3, due to the stable conditions of the plant, a SGTR event is considered to be highly unlikely. Also, the number of PORVs required for heat removal to mitigate a SGTR event is less in Mode 3. While the probability of a SGTR event is not affected by NO SIGNIFICANT HAZARDS CONSIDERATION 58 5/15/97

i IMXISTRY TRAVELERS APPLICABli TO SECTION 3.4 1 a 1 4

        -g TRAVELER #                  STATUS                                         DIFFERENCE #           COMENTS                                      -

TSTF 26 Incorporated 3.4 32 - Approved by NRC. TSTF 27. Rev. [ Incorporated 3.4 33 Apmer/lyA/Ac. ((d j-T3TF 28 Incorporated 3.4 22 Approved b.y NRC. [. 13TF 54, Rev. 1 Incorporated NA d$s*Ob3/cdhnly. g y,4 4 fj d. TSTF 60 Incorporated 3.4 15 Approved by NRC. 3 TSTF 61 Not Incorporated NA Ninor change that is adequately '. y addressed in the Bases. R> TSTF 87, Rev. f Incorporated 3.4 31 Affrsve/ lyM40 - /'g* 3,49 f; TSTF93,Rev.$ Incorporated 3.4 17 N M p % M /or N a'way in OL L' L3 8 74,4-3 Amendment No. 105. _g?.f x3 '- 1 Not Incorporated TSTF 94;gev, l NA Retained current TS. g.-y,4g- , k

              ""1%/=./                 :u,~, .M                                             3..;-36                        g 24/_/ gy.2 g
                                                                                                                                                          ~

TSTF 108. Rev. 1 Not Incorporated NA LCO 3.4.19 does not apply. TSTF 113 Rev. [horporated 3.4 39 d ;3',4,//,-/[$ Rf,//-3

                                                                                                                                       ._S ;,d-  3/j (

2 TSTF 114 Incorporated NA Approved by NRC: Bases 3.4.7 ' changes only. l TSTF 116. Rev. [Yncorporated 3.4 36 j9;T4/2-2 37,y g// i TSTF 136 Incorporated NA B( E E1 ald U .9 changes L only. , 7g y44 'g TSTF 137 Incorporated NA B[ds N Nch h nly. nt y,4, g . TSTF 138 Not Incorporated NA Inconsistent with RCS loops requirements ol' ITS 3.4.5 and 3.4.6. 1" TSTF 151gev. / Incorporated NA Bases 3.4.11 changes only.7#-S.4rpf i TSTF 153 Incorporated 3.4 01 A/y'**VM k NAC- 74-%fM TSTF 162 y Incorporated NA EsN.9[chInhI>'only. '/5-34-gh

Incorporated g g , 3.4 45, See also $s3'.T-[a*nd 3.520. NY I dorpora W _ 3d p ,o g f fj ;

   &         -MO$N-                     Incorporated                                        3.4 10           DCPP only.[jyrpelJyNho,.7p_g,4      q[

j' WOG87,gev.2. Incorporated 3.4 47 $If;//-f- n' y.f,oe MARK UP OF WOG STS REV 1 (NUREG 1431) 5/15/97 l'

Pressurizer PORVs B 3.4.11 ( BASES (continued) SURVEILLANCE SR 3.4.11.1 REQUIREMENTS Block ralve cycling verifies that the valve (s) can be el ... ..,.w_. OA A.

                                                                                                                                                                                                                                                     ,,,..1             -,h._

ss r Frequency of 92 days is the ASE Code, on XI Md/-+ l

f. . w$M. .km hi n,.k u. gl. u. a. d. e ,nimmmJ &_ '

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                                                                                        --__ .                                              - - - -.- ..     - * -.wi- - ^e 's6 wn-              vi      r.-
                                                                                                                                                                                                         ..         run'"s
                                                                                                                                                                                                                    =              bu
                                                         - - . . .                _m       .             ,_ _                 _ _ _ _ , _ ,__ _,                       m_                  .      >__2          . _ , __
                                                      - vr uwww. . u ..                                  u...,                s.w . .x. . .. . v.                        ..             nyo n .o m.                     n .. -

falf111s a zu. p4: Gen-/

                             %f, fg                      _. a.

s not required tU A .M.// .*2 g g,.4.g. fwNote I modifies this SR by stating that

                              /'             ffh'            he equired Acti                                                                                                                                                                       W.!w Q .v.4. ll-+

p _. pent -tie bloe valve in Q24/1-+ (p,

                                                                                                                                                             . ,L hts'              c             i-/Gn incmrer Ne SR 3.4.11.2                                                                                            1-f.tk                      en t.tnf.rolable fea) $wn 4),e Ac.r,rrnee +Ae foRV fr alredy SR 3.4.11.2 requires a complete cyc "Y                                                                                                                       Operating a PORV through one complete cycle ensures that the PORV can be                                                                                                                                                        )

manually actuated for sitigation of an SGTR. 'h: F-^',;;r. y of h/c-3.f-ooy

                                                          =       :--th:             ': based cr. : tuaic=' ref;alir.;; cycl ur.
                                                            ..             +.a ... +4- - 0 oe rahay operience har s'Asw                                                                                             +A ,d   + +hereir.dustry
                      *12rk1%pmheogen*fvalver y                                                                                           un //y air H,asif,-vpfi2 Men,f*,-fu
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                                                                                                                                                                                             . . .' o .c r' m 6 :._

(continued) MARK UP OF NUREG 1431 BASES B 3.4 61 5/15/97

CHANGE NUMBER JUSTIFfCATION_ jk but would limit the exception to prior to entering MODE 2. This change is consistent with traveler RCO.- 7JT/'-JJP, $ f, f,//-a. 3.4 36 SR 3.4.13.1 and ACTIONS for LC0 3.4.15 are revised with the addition of l a note per traveler TSTF 116,R^".1. The note addresses the concern / that an RCS water inventory balance cannot be meaningfully performed ) unless the unit is operating at or near steady state conditions. The Note added to the surveillance provides an exception for operation at less than steady state conditions. The RCS water inventory balance will only be allowed to be deferred for 12 hours after re establishing steady state conditions. 3.4 37 The primary to secondary leakage limits are revised per Callaway OL Amendment No.116 dated October 1,1996. 3.4 38 tent with traveler TSTF 105, the details on by which the RCS - is verified are moved f 3.4.1.4 to the Bases. Moving this informatio ows the use of precision heat balances, elbow taps. r acc thods in order to perform this verificat nd is consistent with the N - ev. 1 philo moving clarifying information and descriptive s out ha TS to the Bases _f f y / , ./ , C , p , ,f,' , ' f) T,f,/-/ nver. tion Cws/wo-i.ron'g~ible 7 Enc /s.rarat $8 3.4 39 The shutdown regtfirements of ITS .4.11 would requi e the plant to reduce Tm to < 500*F within 12 hours, rather than go to MODE 4, to address the concern of entering [COMS] LCO 3.4.12 Applicability with inoperable PORVs. For consistency, the shutdown requirements of ITS 3.4.16 are al o revised to allow 12 hours to rAedu e T_ to < 500*F. This change is c nsistent with traveler TSTF

                                                                                    " 113/.ZNAEgt"gA-f] 8 84 //'5
                                                                                                ^

7"f7F'- 2PR #

3.4 40 Consistent with traveler "CC ^^,Vthe Note to SR 3.4.1.4 would be 8 f,f, /-2 l modified to provide additional time to perform an RCS precision flow rate measurement. The time allowed would be changed from 24 hours to 7 days. This change is acceptable because other indication of RCS flow is available (SR 3.4.1.3, RCS total flow meters) and additional time normally would be required to establish plant conditions suitable for the precision heat balance. Since this parameter does not normally change significantly and the flow meters can be used in the interim, there'is no need to perform this SR within the 24 hour period specified in NUREG 1431 Rev.1. The 7 day period provides sufficient time to establish steady state plant thermohydraulic conditions and obtain equilibrium xenon. In addition, the THERMAL POWER specified in the Note would be changed from the generic value in brackets (90% RTP) to 95% RTP. This change is acceptable because it specifies a power level l in better agreement with current operating procedures for performing a ! precision heat balance. Current TS do not specify a power level for j this measurement. 3.4 41 Not applicable to Callaway. See Conversion Comparison Table (Enclosure 6B). JUSTIFICATION FOR DIFFERENCES TS 6 5/15/97

i l I ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION WO: Q 3.4.11-3 APPLICABILITY: CA, CP, DC, WC REQUEST: Change 4-05 LS 31 and Difference 3.4-39 Comment: TSTF-113 (presently Rev. 4) has not yet been app.oved by the NRC staff. l FLOG RESPONSE: TSTF-113 Rev. 4 revises the shutdown requirements of ITS 3.4.11 to allow the plant to reduce Toy to <500 F within 12 hours, rather than MODE 4, to address the concern of entering LCO 3.4.12 Applicability with onc or more inoperable PORVs. The shutdown requirements of ITS 3.4.16 are also revised, for consistency, to allow 12 hours to reduce T., to < 500 F. ITS 3.4.11 Condition B and C Bases changes have been made to the Callaway submittal to reflect Rev. 4 of the traveler; no changes are required for any other plants'submittals. The FLOG continues to pursue the changes proposed by this traveler. ATTACHED PAGES: Attachment 10, CTS 3/4.4 - ITS 3.4 Enclosure 5A, Traveler Status Sheet I Enclosure 58, page B 3.4-58 l l l l l I l l l l

IWUSTRY TRAVELERS ~ APPLICABLE TO SECTION 3.4 . TRAVELER # STAT 1]S DIFFERENCE # COMENTS - TSTF 26 Incorporated 3.4 32 Approved by NRC. TSTF 27 Rev. [ Incorporated 3.4 33 Ajpmell yA/4C, [M g I TSTF 28 Incorporated 3.4 22 Approved by NRC. [ TSTF 54 Rev.1 Incorporated NA Ns*Ob37chIhnly. ry y4 4 1 TSTF 60 Incorporated 3.4 15 Approved by EC. j TSTF 61 Not Incorporated NA Minor change that is adequately :. y addressed in the Bases. 4 TSTF 87, Rev. f Incorporated 3.4 31 A/ /rW[lyNAC - 7ps4y [ Incorporated 3.4 17 N M ap3 M M /or E ia'way in OL TSTF 93. v Rev.)3 & 2,+,9-3 Amendment No.105. _ gyp 23 TSTF94;gev,l Not Incorporated NA Retained current TS. g-y.45- KI

      .;:- 1"/= /            lc.c.;,, r. M                             3."-36                            g 24/_./ gy.f gjj TSTF 108. Rev.1         Not Ij)corporated                           NA            LCO 3.4.19 does not apply.

8 h,/(,-/[$f[//-S , W,4

TSTF 113. Rev. [Yncorporated_ - 3.4 39 3//- TSTF 114 Incorporated NA Approved by EC: Bases 3.4.7 . , changes only, f, l u! TSTF116,Rev.d~Yncorporated [S4/F-Q %fny 3.4 36 a// fl TSTF 136 Incorporated NA B( E 43 #Ma U .E changes i: only. ,, -rg-fA.y9} TSTF 137 Incorporated NA BMMkchf(( Inly. ng sg. p ' TSTF 138 Not Incorporated NA Inconsistent with RCS loops requirements of ITS 3.4.5 and .- 3.4.6. TSTF 151gev. / Incorporated NA Bases 3.4.11 changes only. 74'Sh pf [N TSTF 153 Incorporated 3.4 01 [//*NVM k A/AC- TE-J,fM TSTF 162 y Incorporated NA Es3.EchInN>'only. 'TAL-34-g Incorporated M d ".3, 3.4 45, See also hs3'.T'-[and 3 I~#

20. D i 1;w.1c ~aET ~

      - ..                    Incorporated                             3.4 35                          g y,4g-p                      --co  a   ->      1 gip    ??ON                    Incorporated                             3.4 10           DCPP only./jyryw/ lyn #d,yg_y,4 gg !r$

WOG 87, /cv. :2. Incorporated 3.4 47 $ 7.fi//- f 7 " 7.f e? MARK UP OF WOG STS REV 1 (NUREG 1431) 5/15/97 m-c-,,- w,-- r,, , , - - -,m ,- -ee- ~ 3-,

Pressurizer PORVs B 3.4.11 1 BASES l ACTIONS B.1. B.2. and B.3

 ~(continued)

If one er-two PORVt is inoperable - and not capable of being l manually cycledg it must be either restored or isolated by l closing the associated block valve and removing the power to the l associated block valve. The Completion Times of 1 hour are reasonable, based on challenges to the PORVs during this time period, and provide the operator adequate time to correct the situation. If the inoperable valve cannot be restered to OPERABE status, it must be isolated within the specified time. Because there is at least one PORV that renains OPERABLE, an additional 72 hours is provided to restore the inoperable PORV to OPERABLE status. If the PORV cannot be restored within this ! additional time, tM pi;at ;;t k tre.g,t t; ; OC in J,id tM LOO d= ret y, n 7;; ired by Condition D,gk8Uh3 ) muhle$rayM-E e? /ea.r/- c.1 and c.2 fM 3-w WM4 7~Ti < 90 A r re[uirefly j If one block valve is inoperable, then it is necessary to either restore the block valve to OPERABLE status within the Completion Time of 1 hour or place the associated PORY in manual control. The prime importance for the capability to close the block valve is to isolate a stuck open PORV. Therefore, if the block valve cannot be restored to OPERABLE status within 1 hour, the Required Action is to place the PORV in manual control to preclude its automatic opening for an overpressure event and to avoid the potential for a stuck open PORV at a time that the block valve is inoperable. The Completion Time of 1 hour is reasonable, based on the small potential for challenges to the system during this time period, and provides the operator time to correct the situation. Because at least one PORY remains OPERABLE, the operator is permitted a Completion Time of 72 hours to restore the inoperable block valve to OPERABLE status. The time allowed to restore the block valve is based upon the Completion Time for restoring an inoperable PORY in Condition B, since the PORVs are not .

                                 - capable of mitigating an everpressuee event when ple;ed in ;;aal ;eatrel.         .,  ' - . . . . -     ,4,    -

If the block valve is restored within the Completion Time of 72 hours, tM p;;;r will k r;;tered ;ad the PORV r;;tered t; 0""J"L: ;tets;. i< - u .~3, If it cannot be restored within this additional time. ( tM pi;at at k breu#,t t; ; =C in J,id tk LOO dec; ret

                   ;pply, ;; required b7tondition D,g                                       d2 3.t.//-3 ZWfeA?7-ffj.cp                                                         0#*Ul4 (continued)

MARK UP OF NUREG 1431 BASES B 3.4 58 5/15/97

i ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.11-4 APPLICABILITY: CA, CP, DC, WC REQUEST: Change 4-09 LS-36, Difference 3.4-47, Change 3-04 and Difference 3.4-31 Comment: WOG-87 has not yet become a TSTF. FLOG RESPONSE: As discussed during a telecon with NRC Staff on July 30,1998, the above references to DOC 3-04 and JFD 3.4-31 apply to NRC-approved traveler TSTF-87 and were not intended to be questioned here. Additional changes have recently been added per Revision 2 of WOG-87 and are included in the attached pages below. The addition of the Note to the block valve Action Statement is considered to be administrative in nature as it reflects current plant practice. WOG-87, Revision 2, has been approved by the TSTF group and is expected to be submitted to the NRC expeditiously. Given the nature of the Notes added to the PORV block valve Required Actions and Surveillance Requirement, the FLOG continues to pursue the changes proposed by this traveler. ATTACHED PAGES: Attachment 10, CTS 3/4.4 - ITS 3.4 Enclosure 2, page 3/4 4-10 Enclosure 3A, page 7 Enclosure 38, page 5 Enclosure 4, pages 60 and 61 Enclosure SA, Traveler Status Sheet and pages 3.4-21 thru 3.4-23 Enclosure SB, pages B 3.4-58, B 3.4-60, and B 3.4-61 Enclosure 6A, page 7 Enclosure 68, page 6 i

REACT 02 COOLANT SYSTEM 3/4.4.4 REllEr VALVES

      'TMITING CONDITION FOR OPERATION 3.4.4 Both' power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.* Ndc: refonk CondrHon en4ry rr aIlowel & enel fokV ACTION: 4 -p l- U

a. With one or both PGRV(s) inoperab an)c ab/*oflei r ::fr,: :_st nunua//, h: l, eyelet S;e ,O2'M within I hour either restore the PORV(s) to OPERABLE status or close the associated block valve (s)with Dower maintaineri 3 tn the hinck valve

                                                  .5MU'00 Z within the foll ino 6 hours,                                                               a av(no+ e+** Vbat bergro*wa/ly cycle),%                                   . 05::::ndee 7"1         6<s

b. With one PORV inop6rable du: .. :::::. ..hrr _ 17: :::t 4--42-LS 1:&;;, within 1 hour either restore the PORY to OPERABLE status, or close its associated block valve and remove power from the block '

g valve- restore the PORY to OPERABLE status within th; f;1 h.dr.; sf.-a.7-A1 72 hours orfbe in HOT STANDBY within the next 6 hours an " "a-

                                       - E"J' TD0"" within the following 6 hours.                                                          N,Inc G -/,          < fd/*[ pOf_ cy

c. WithbothPORV[pinoperable.de:y an,(no)-cgalk t: ::::. nua/k Oth:r c rof::::::fr:

                                                                                                                                         .h_yc     Jery:

t **

s y, pp/,5

         #                                    '- W - within status                                ortcloseI hour either restore at le t one PORV to OPERABLE its / associated block valv ' and remove power from and be in HOT STANDBY wi           in t'.; r,;;; 6      urs and        s/.-67-M

, the u;T SM"T block% valv@ wi thin the following 6 hou a gf, // 4 sf gf LJ

d. reduce ~1Q h <SDe*f p 4-p q-.L. S With one or both 6 lock valves inoperab e, within I ho_ur res re the block valve (s) to OPERABLE status or p ociated PORV(s) in manual control. Restore at least one block valve to OPERABLE status w:ithin the next hour if both' valves are inoperr.ble; restore any

         "                               remaining inoperable block valve to OPERABLE status within 72 hours; otherwise,Vbe in at least HOT STANDBY within the next 6 hours and .k- 6df--45 HOT :""TOO"" # thin the following 6 hours.                                                             repea "         %g< SMadf

e. The. pec tsions of Specification 3.0.4 are not applicable.

SURVEILLANCE REOUIREME V,8*fte[y 4.4.4.1 '-

$ ti:r t:$.he requirements of Specification 4.0.5, :::r m-

_m_,. m m__.__.a w. e . e , r ..i....______*e _..u. u.. --_<--. .. _< . b8bd6 E.E c F :liiEElh W (I E litiitIir W 5 tiB E 7 fl.e % r A e'd Wf <#, ~ cle o$ each 4/V fr onlyregaintol da Jefer-for.rs,collol"/%bE.f lad l2, = -Gr LS 4.4.4.2 q [e4e cy. Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one coeplete cycle of full travel unless the block valve is closed in'o'rder to meet ' " - d - " " c ' .* C T : C" b . ;; : .

                                                   '                               in Speci fication 3.4.4. DA/y' rehfers/ f ie f er$ormeslIn Nik.f lnov/2. QT-LS ACT.I2M/f                                                                                  4 gejLLJ "With all RCS cold leg temperatures above 368'F.
      .LLAVAY - UNIT 1                                                                                    3/4 4-10        Amendment No.83 cond-inae eerhrn-h'on ac-lirlf*er ard                                                                                    ,               v                      -

- - A c r.roiTei doae na+ ajyty wAan R,a Llock nt;c(r.) ir inofe<>Jle to/*17 +-0 9-LS

  - of a reta/+ of u^f ging w;% ActroAlf A or c ,                                          /                                                                          A .1%Il-+

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

l CHANGE NlkBER HSHC DESCRIPTION steam dump] valves. This change is consistent with traveler TSTF-113. Rev.' 3. 4 06 LS 32 Not applicable to Callaway. See Conversion Comparison Table (Enclosure 3B). 4 07- LS 33 Not applicable to Callaway. See Conversion Comparison Table (Enclosure 38). T.51f-:199' l 4 08 LS 34 Consistent with traveler E 50,Vthe requirement to O M,//d I perform the 92 day surveillance of the pressurizer PORV l block valves and the 18 month surveillance of the i pressurizer PORVS (i.e., perform one complete cycle of each valve) is revised to indicate that the surveillance is only required to be performed in MODES 1 and 2. This !

is consistent with the reconnendations of Generic Letter 90 06, Resolution of Generic Issue 70, ' Power Operated Relief Valve and Block Valve Reliability ' and Generic Issue 94, ' Additional Low Temperature Overpressure Protection for Light Water Reactors,' Pursuant to 10CFR50.54(f)," June 25, 1990.

l 4 09 LS 36 The requirement to perform the 92 day surveillance of the pressurizer PORV block valves (i.e., perform one complete cycle of each block valve) is revised such that it is not required if the block valve is closed to meet ACTION a of current TS LCO 3.4.4. This change is acceptable because no credit is taken for the automatic actuation of the PORV in Modes 1, 2, or 3. Credit is taken for manual operation of the PORVs during the Steam Generator Tube Rupture (SGTR) accident. However, the capability to manually cycle the PORVs will be unaffected by this change. This change will not affect the ability of the block valve to open, if closed to meet ACTION a, in the mitigation of an SGTR. Deferral of the block valve cycling surveillance will not diminish the design capability of the block valve to open against differential pressures that would present , after an SGTR since the block valves are capable of l opening against 2485 psig, the safety valve lift pressure. ; h g Q Lereas pressurizer pressure decreases after an SGTR. g g g g his change is consistent with traveler WOG-87. 5 01 A This change moves the steam generator tube surveillances to ITS SR 3.4.13.2 and the Administrative Controls Sections 5.5.9, SG Tube Surveillance Program, and 5.6.10, 1 SG Tube Inspection Report. L I DESCRIPTION OF CHANGES TO CURRENT TS 7 5/15/97 L L ;

INSERT 3A-7 Q 3.4.114 l l In addition, a Note is added to ACTION [d] stating that it does not apply when the block valve (s) is inoperable solely as a result of its power being removed per ACTIONS (b or c) as a l result of an inoperable PORV(s). In this scenario ACTION (b or c) is entered as a result of an inoperable PORV(s). If one PORV were inoperable and incapable of being manually cycled i (per the change discussed under DOC 4-02-LS-6), ACTION [b] would be entered at time zero, to. ACTION (b) would close the associated block valve and remove its power within time to + 1 hour. If, as a result of block valve power removal per ACTION (b), ACTION [d] were then entered, ACTION (d] would require the associated PORV to be placed in manual control within time to + 2 hours. However, the reason for originally entering ACTION (b] is that the associated PORV is inoperable and can't be manually cycled, thus there is nothing to be gained by placing the PORV in manual control. The PORV inoperability may be such that the PORV can't be placed in manual control (e.g., blown control power fuse), in which case neither this portion ) of ACTION [d] nor block valve restoration can be met. In addition, the portion of ACTION [d] l requiring block valve restoration can't be satisfied as long E:s power is removed from the block ..

valve. Restoring the PORV to OPERABLE status within time to + 72 hours allows the plant to exit ACTION (b). If power were not restored to the block valve at this time, the new Note on ACTION [d] would have no standing and ACTION [d] would be entered. Similar conclusions can be drawn for the relationship between ACTIONS (c and d]. If ACTION (c)is the original

, ACTION entered, there is nothing to be gained by placing both PORVs in manual control and the block valves can't be restored with their power removed. With ACTION [d] not satisfied, the plant must go to MODE 3, but ACTION [c] would have already had the plant in MODE 3 two hours earlier. Therefore, there is no compensatory action associated with cascading to the ) block valve ACTION (d] when the sole inoperability is with the PORV(s). l l l 9

CONVERSION COMPARISON TABLE - CURRENT TS 3/4.4 Page 5 of 13 TECH SPEC CHANGE APPLICABILITY DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY NUMER DESCRIPTION The shutdown requirements of CTS LCO 3.4.4 would require the Yes Yes Yes Yes 4-05 LS-31 plant to reduce T,.,to < 500*F within 12 hours, rather than go to MODE 4. to address the concern of entering [COMS] LCO Applicability with inoperable PORVs. For consistency, the shutdown requirements of CTS LCO [3.4.8] would be similarly revised. This change provides a 72 hour completion time to restore an Yes No - Already part No - Already part No - Already part 4 06 LS-32 inoperable block valve. with the PORY placed in manual of current TS. of current TS. of current TS. control mode. The current TS requires the block valve to be restored within one hour, or remove power from the solenoid. This change provides a two hour completion time for Yes No - Already part No - Already part No - Already part 4-07 , LS-33 restoring an inoperable block valve when more than one block of current TS. of current TS. of current TS. valve is inoperable, and 72 hours to restore the remaining - valves. The current TS requires the block valve to be restored within one hour for one or more valves inoperable. Consistent with traveler = ".dhNe'quYr~eMb perform Yes Yes Yes Yes N //~2 4 08 LS-34 the PORY and block valve cycling surveillances is revised such that the surveillances are only required to be performed in MODES I and 2. Consistent with traveler WOG-87, the requirement to perform Yes Yes Yes Yes 4-09 LS-36 the 92 day surveillance of the pressurizer PORV t, lock valves (i.e.. perform one complete cycle of each block valve) is revised such that it is not required if the block valve is closed to meet ACTION a of current TS LCO 9 3.4.4. T a///1%gA/ap /r $4Y Y8 [ ef' fb j&eved e4 .ra /a/7 f),2f/j_4 I 5 01 This change moves the Steam Generator Tube Surve11Ya"ncesYes to "o N Yame as N b - Ye'sN' Yes A ITS SR 3.4.13.2 and the Adninistrative Controls Sections change 1-14-A for 5.5.9 and 5.6.10. CTS Section 3/4.0. CTS LCO 3.4.5 is deleted. Steam Generator operability Yes No - CPSES does not Yes Yes 5-02 A requirements in MODES 1-4 are specified in the RCS loop and have this i

leakage spectf1 cations. spectfication. Clarification to remove potential interpretation problems Yes No - Same as CPSES Yes Yes 5 03 A related to probe orientation versus entry point. change 1-15 for CTS Section 3/4.0. CONVERSION COMPARISON TABLE - CURRENT TS 5/15/97

IV. SPECIFIC NO SIGNIFlCANT HAZARDS CONSIDERATIONS

   ,c.   +

10 CFR 50.92 EVALUATION FOR TECHNICAL CHANGES THAT IMPOSE LESS RESTRICTIVE i REQUIREMENTS WITHIN THE TECHNICAL SPECIFICATIONS Consistent with traveler WOG 87, the requirement to perform the 92 day surveillance of the pressurizer PORV block valves (i.e., perform one complete cycle of each block valve) is ACTION revised such that it4 is not a of requjred if the block valvg js/ closed to meet-/s Acco4/ d b,oreverr/currbnt en y roleTS Lgue.4 L .N/4 a Nefef.emueinyereita /gy(s) NIIA J oc c .under Acr.r This proposed TS change as been evaluated and it has been determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92(c) as quoted below:

                       "The Comission may make a final determination, pursuant to the procedures in 50.91, that a proposed amendment to an operating license for a facility licensed under 50.21(b) or 50.22 or for a testing facility involves no

. significant hazards consideration, if' operation of the facility in accordance with the proposed amendment would not: 1 l 1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or v. l 2. Create the possibility of a new or different kind of i.ccident from any l accident previously evaluated; or e 3. Involve a significant reduction in a margin of safety." l The following evaluation is provided for the three categories of the significant hazards consideration standards: ,

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

The proposed change adds a relaxation to the surveillance associated with the pressurizer PORV block valves. The quarterly valve cycling will no longer be required if the block valve is closed per any ACTION of the LCO. No credit is taken for the automatic actuation of the PORV in Modes 1, 2. or 3. Credit is taken for manual operation of the PORVs during the Steam Generator Tube l Rupture (SGTR) accident. However, the capability to manually cycle the PORVs will be unaffected by this change. This change will not affect the ability of the block valve to open. if closed to meet ACTION a. in the mitigation of an SGTR. Deferral of the block valve cycling surveillance will not diminish the 4 design capability of the block valve to open against differential pressures that would be present after an SGTR since the block valves are capable of l opening against 2485 psig, the safety valve lift pressure, whereas pressurizer NO SIGNIFICANT HAZARDS CONSIDERATION 60 5/15/97 4 1

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

IV. SPECIFIC NO SIGNIF1 CANT HAZARDS CONSIDERATION NSHC LS 36 (continued) pressure decreases after an SGTR [as shown in FSAR Figure 15.6-3A]. The lack of quarterly block valve cycling, which could extend to a complete cycle since ACTION a allows continued operation with the block valves closed, does not decrease the likelihood of successful pressurizer relief since power remains available to the block valve motor operator (s) and the surveillance frequency for the PORVs can be as long as 18 months (tested during each cold shutdown per the IST plan). Quarterly cycling could make PORV seat leakage worse; if the block valve were to subsequently be unable to close, this surveillance could unnecessarily challenge RCS and PRT integrity. Therefore, the proposed change does not involve a significant increase in t probability or consequences of an accident previously evaluated. g fgj_.4

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

The only accidents that are potentially associated with this proposed change are those related to a loss of pressurizer relief function. This change does not introduce any new overpressure accidents and the existing analyses remain valid. Thus, the proposed change does not create the possibility of a new or different kind of accident from those previously evaluated.

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

The proposed change does not affect the acceptance criteria for any analyzed event. The automatic actuation of the PORVs is not credited in the accident analyses for Modes 1, 2, or 3. The PORVs will remain capable of being

             ; manually cycled. The margin of safety established by the LCOs also remains unchanged. Thus there is no reduction in the margin of safety from that previously established.                                       ,

NOSIGNIFICANTHAZARDSCONSIDERATIONDETERMIhTION Based on the above evaluation, it is concluded that the activities associated with NSHC LS-36" resulting from the conversion to the improved TS format satisfy the no significant hazards consideration standards of 10 CFR 50.92(c): and accordingly, a no significant hazards consideration finding is justified. N0 SIGNIFICANT HAZARDS CONSIDERATION 61 5/15/97

INSERT LS-36 Q 3.4.11-4 l I l The addition of the Note to ACTION (d) stating that it does not apply when the block valve (s) is l inoperable solely as a result of power being removed per ACTIONS [b or c) as a result of an inoperable PORV(s) eliminates operator distraction caused by the required performance of . activities with no safety benefit. This change has no effect on the successful mitigation of an SGTR since the initial premise is that the PORV(s) is unavailable. Elimination of operator actions that have no safety benefit result in an overall benefit to plant safety. i j i l l i

 ,_           m                                                                                                                           -,--

IiETRY TRAVELERS ~ APPLICABLI TO SECTION 3.4 f TRAVELER # STATUS DIFFERENCE # CO> TENTS - TSTF 26 Incorporated 3.4 32 Approved by RC. TSTF 27 Rev. [ Incorporated 3.4 33 Ajpme//fA/4d, ((f; jg b TSTF 28 Incorporated 3.4 22 Approved b.y RC. { TSTF 54, Rev. 1 Incorporated NA S$s*Ob37chIhnly.7E s4p 1 TSTF 60 Incorporated 3.4 15 Approved by NRC. Q TSTF 61 Not Incorporated NA Ninor change that is adequately $

                                             ,                                                          addressed in the Bases.                                                       E TSTF 87, Rev. f Incorporated                                                  3.4 31           AffrWM yl NAO - 7p.3 4y                                                    !

TSTF93,Rev$ Incorporated 3.4 17 N M p % M /or M ia'way in OL . I L3 4 7,f,4-3 Amendment No. 105. g?.f ;;3 p, t TSTF94;gev,l Not Incorporated NA Retained current TS. 7g y.Mg-

          = mJ=. !                                :.u. ,~. .w                          3.eas                                            g gg/-/ gr 2 ::/                               9j TSTF 108, Rev.1                         Not Incorporated                        NA            LCO 3.4.19 does not apply.                                                 !      !

4 - - 13TF 113, Rev. (fTncorporated 3.4 39 d ;T.4,/4-/[$2f,//-3 ~7Z' 7,N/j l ! r1 TSTF 114 Incorporated NA Approved by NRC: Bases 3.4.7 ' l changes only.

o. ,

TSTF 116, Rev. d~ Morporated 3.4 36 j9jf 4/F ,'2 -r'"

                                                                                                                                                                       ;, f- g/l !3 i 13TF 136                                 Incorporated                           NA            BfE43 [dd U.Ichanges                                                       i only.                        ,,

rg-y44}

TSTF 137 Incorporated NA B[MMdch7hnly. -rX' JA W '

TSTF 138 Not Incorporated NA Inconsistent with RCS loops requirements of ITS 3.4.5 and 3.4.6. J-w TSTF 151gev. / Incorporated NA Bases 3.4.11 changes only.74-S.Mfk T5TF-153 Incorporated 3.4 01 A/fr*VM k A/AC J,fM j TSTF 162 y Incorporated NA Es 3N.9/cEnhI>'only.'7E-34-gh Incorporated g g3 3. 3.4 45, Seealsoks3'.T'-[a$d3.420['I i D-uxe ~avr Incorporated 3.4 35 g y,4ff-p --r-o gf jj_ ( l6 '-MO59 __ Incorporated 3.4 10 DCPP only./)pr,w/j y Ngd,-7p_y,g [l

                                                                             ~s           x        ~

WOG 87, gev. :2. Incorporated 3.4 47 f://-f, n' -'.f mr 4

                                                                           -         A NARK UP OF WOG ST3 REV 1 (NUREG 1431)                                                                                                     5/15/97                   t b

                                                        --                 NoTG - - - --

faguiral Acknr do ,,,+ yyy l 1

     ~~'                                         tv A s n 4 lo c k W /ve fe /n en//c Pressur%rPORVs               3 4 11 j                                              solely as a resul+rf coglyt3 wi+f f uired Acnr 8..:2 o,- E.s.

k, ;,-) ACTIONS (continued) f _3 CONDITION REQUIRED' ACTION COMPLETION TIME G+ 2.+41 C. One block valve' C.1 Place associated PORV in 1 hour inoperable. manual control. M C.2 Restore block valve to 72 hours i OPERABLE status. l D. Required Action and - 4~- - e- - - M J3.4 associated Completion o.- - 1 Time of Condition A. B, or C not met.

                                                     . 2 :" '        - ".
                                                                                                                                             )

5 D.ig Be in H0DE 3. 6 hours l M

D.El "; ir = 4. 12 hours < 3.4@ E. ,Two er tt.re; PORVs : ~1- - - - x - s3.4-39c inoperable and not ...:eu ..em. o - : .o .-- L B PS = capable of being , manually cycled. E , 1 E.13 Close associated block I hour i valves. M E.Eg Remove power from I hour associated block valves. M

    ;Q7.1                                                                  -

(continued) MARK UP OF WOG STS REV 1 (NUREG 1431) 3.4 21 5/15/97

                                                              ~
                                                   -- - yp y g                 --

AoguTreal Ac h f e/, d g/y Pres rizer PORVs i - when Llock n/ve is inyereAle 3.4.11 so fel y or a re ru/+ o f conyly im k'ff4 AeguiredAc/ rov f.ao , dss. ACTIONS (continued) - - CONDITION REQUIRED ACTION COMPLETION TIME E. (continued) E.S{ Be in H00E 3. 6 hours M E.4[ a in = 4. M 12 hours 3.4 394 e . s.#m F. More than one block ".1 F Place associated PORVs in 1 hour Q Sd. II-+ valve inoperable. manual control. M F.2 Restore one block valve to 2 hours OPERABLE status, i' three b1 x k velv;; er; ir.egrebi;. ..B PS M T.3 ";;ter; is..inir- bleek 72 b ar; 'B PS1 e... velv;';) to 0"C"X'l st;te;. G. , Required Action and E.>- - > . -- -

                                                                                                   ~ +                  .'3;4 39-associated Completion                  -

7 - . a , a:" m - Time of Condition F not - met. M ' G.tg Be in H0DE 3. 6 hours M G.2[ "; in ".00C v.- M 12 hours 3.4 39 Mi$ MARK UP OF WOG STS REV 1 (NUREG 1431) 3.4 22 5/15/97

t i. i Pressurizer PORVs - 3.4.11 4 l l 4 (' SURVEILLANCE REQUIREENTS SURVEILLANCE FREQUENCY , i

                                                                                                                                                            %                                                     I SR 3.4.11.1                                                   -

No -- - -- w - Not required met M with :.3.4 -47J l block valve osed in accordance with the l Required ionsof 0; .ditier, "" ;r " -/4 ff/_ Co. .

                                                                                                                                                                                        ,J T      a f2,f1,/---^E+

j E .. , 3.4 35

                                                                                                                                                 ~                  v

Perform a complete cycle of each block 92 days

                                                                                                                                                                 $ 3.+.ll 4' -7,2 ..                          u x                -                         ~                  -
                                                                                                                                                                . _m                                ,.m SR 3.4.11.2                                                                                                                            "'IM S*

Dw vice 13.4 35'. g Perform a complete cycle of each PORV. = _ ./.:.. aBr 1 Z 3.4.11.3 8.PS;

                                                       ";rfere ; glet; ;yci; ei ;;;t, .;1;..eid eir                                                             ifHeenths-
                                  ,                    centr;l velv; ;,,d ;t.;;% velv; ;- tt.; eir ec;niuter; ir, "0'".' ;.7,trel ;j;t;;;.

e. , , , , , .,__,,.._.,_m < , _ _ . . .._,..__ ___ _ __<,_ _,

e#W) ad e 'T . e e . 'T vb4 5Eg I ws \ W d w3 3%5 p s wbr% vuIvbd w3 h bufww3E v3 'w sJ' rry- y;wei ;d ive As zwir;g [w';r E r;;;. 4 ,

          ,  'a#
,        .< ., t .
       *y.          ,

o l MARK UP OF WOG STS REV 1 (NUREG 1431) 3.4 23 5/15/97

Pressurizer PORVs B 3.4.11 i i BASES

;         ACTIONS                  B.1. B.2. and B.3 (continued)

If one e@ PORVs is inoperable end not capable of being 4 manually cycledE it must be either restored or isolated by closing the associated block valve and removing the power to the associated block valve. The Completion Times of 1 hour are i reasonable, based on challenges to the PORVs during this time

,                                  period, and provide the operator adequate time to correct the situation. If the inoperable valve cannot be restored to 4

OPERABE status, it must be isolated within the specified time, i Because there is at least one PORV that remains OPERABE, an , additional 72 hours is provided to restore the inoperable PORV to OPERABE status. If the PORV cannot be restored within this l additional time, the plent ;;;t b; bri,ugt te ; = in Jich the LOO di,e; 7.et ; y. n 7;;; ired by Condition D, k8I/h3 , m u O e ],ra C.1 and C.2 f SMM WI'M K <f-f n-l- leu}-S 0 K A f q - l If one block valve is inoperable, then it is necessary to either , restore the block valve to OPERABE status within the Completion Time of 1 hour or place the associated PORV in manual control. l The prime importance for the capability to close the block valve I is to isolate a stuck open PORV. Therefore, if the block valve cannot be restored to OPERABE status within 1 hour, the Required i Action is to place the PORY in manual control to preclude its i automatic opening for an overpressure event and to avoid the potential for a stuck open PORV at a time that the block valve is I ! inoperable. The Completion Time of 1 hour is reasonable, based i on the small potential for challenges to the system during this time period, and provides the operator time to correct the ! situation. Because at least one PORV remains OPERABE, the ! operator is permitted a Completion Time of 72 hours to restore

the inoperable block valve to OPERABE status. The time allowed

! to restore the block valve is based upon the Completion Time for i restoring an inoperable PORV in Condition B, since the PORVs are d . - capable of mitigating an n;rpra;;r; event when pieced in ;;r.nl catrel. j "- - If the block valve is restored within the Completion f Time of 72 hours, th; p;ur dll b; ruter;d end the PORV 4 r;;tered te 0""Of ;teta;. -

                                                                                                  --       - ^                    + +
                                       ~~"

If it cannot be restored within this additional time.

j ( th; pl;nt ;ct b; bres;ht te ; = in 2ich th; LOO de;; not l ;n,1y ;; r;;uired b3"tondition D, 6 4 3.t.//-3 zwdi- g y,fh - 1 _ p ' &2+.//-+} (continuea) -

MARK UP OF NUREG 1431 BASES B 3.4 58 5/15/97 1

INSERT B 3.4-58 Q 3.4.11-4 The Required Actions are modified by a Note stating that the Required Actions do not apply if the sole reason for the block valve being declared inoperable is as a result of power being removed to comply with other Required Actions. In this event, the Required Actions for inoperable PORV(s) (which require the block valve power to be removed once it is closed) are adequate to address the condition. I l I i i i N._, j I l r I

Pressurizer PORVs B 3.4.11 BASES

f. l ACTIONS E.1. E.2. E.3. and E.4 - (continued) to reach the required plant conditions from full power conditions g jc in orderly manner and without challenging plant systems. In

                                              , e- 1                                    MODES                                                                                                       $1f,6e/
                                                                                        -           em OPERABILITY g mew required. See LC0 3.4.12.

e - . PORV J r t-F .1, F.2. ;.2 ' . 2 N 8 Y* *" 'I fIf more than one block valve is inoperable, it s necessary (rtlline) to either restore the block valves within the Comp etion Time of

I hour, or place the associated PORVs in manual control and restore at least one block valve within 2 hour i.74 r;;ter; tt.e $7.6(en-/

                                                                                         . w ir.in; bled ;;1^;; .;itt.in 72 t.wr;. The C                                letion Times are reasonable. . based on the small potential for challenges to the

system during this time and provide the operator time to correct i 3 the situation. . _

l rurgw A2.+ g & 5f.//-+ i , ^ ^ m G.2 m If the Required Actions of Condition F are not met, tt.cr. tt.; pler.t m;t b; ,rw#,t te s = ir. J.id tt.; LOO dec; r.et epply. Te ;di;;; tt.;; ;t;ta;, the plant must be brought to at least N00E 3 within 6 hours and 8 to MOBE--4 within 12 hours. .~

.. . . ... . . , , , .. . .. ,c l
. The allowed Completion Times are reasonable, based on operating experience, to reach the required

. plant conditions from full power conditions in an orderly manner I and without challenging plant systems. In H0 DES '- + .. * $ f.f4,m/

m me+ntowHng . . .':~ . - PORV ILITY may-be B required. See LC0 3.4.12. 4

                                                                                                                                                                    -,A       /r una 9 I

MARX UP OF NUREG 1431 BASES B 3.4 60 5/15/97 4 i

                                                 -. INSERT B 3.4-60                          Q 3.4.11-4 l                  The Required Actions are modified by a Note stating that the Required Actions do not apply if l-                 the sole reason for the block valve being declared inoperable is as a result of power being j                  removed to comply with other Required Actions. In this event, the Required Actions for j                  inoperable PORV(s) (which require the block valve power to be removed once it is closed) are j                  adequate to address the condition.

i.

s p.*' 4 J l

v. ..,

i l

Pressurizer PORVs B 3.4.11 BASES (continued) SURVEILLANCE SR 3.4.11.1 REQUIREMENTS Block ralve cycling verifies that the valve (s) can be

                                                                                                         ,,    _x, C1                   . . .                  . .
                                                                                                                                                                                                                                                                                           , ,s
                                                                                                                                                                                                                                                                                                ~. ,..

m ,m as s or Frequency of 92 days is the ASE Code, on XI S? *//-+

                                                                                                                      ,, . u wie,.9 o,3o.                                                 4_ _,       . . ,_ ._ ._ a  _ ..          2.. _ _ ,. .m. . . .n.

u.. (Ref. 3 E,.). ., . . _... _._ ._.._ __4... ....u

                                                                                           .-r--               "' " "

u4

                                                                                                                                W "'
                                                                                                                                                                'J              "J...u,                    b. s wr6r            .v.-...-

w&64II U. 6s w.o

                                                                                                                                                                                                                                                                                       .m-
                                                                                  "-1 :: n'":f i--- t:=;,""u' c+..';. s,+ W tu tisck = = i:
                                                                                  ====y                        : g=it th: =" to be . .: fu ==:1 n.e :1 r c---'eir

_,,m_..__ Ti= m>m e rut =; tu m! .r.d c_ ,a, tu buc al= :

                                                                                  .--.,......,,m.m..._.m..,,_m                                                                                                                 _ ,._.. ,~ ,, m
                                                                                                  , . c . n,        . .
                                                                                                                                     ..... .                    ,_.....,n....
                                                                                                                                                                                                                                                  .. ~ . _.
                                                                                                .y......
                                                                                                                                                                                                               ,~.

a

                                                                                                ^
                                                                                  .--__--._-1.,u . ____     1 a e.s4    k1. ne.1,_ _ . . .. S. v - up,v.. . ..*6v. . ^b . vi ,
                                                                                                                                                                            - - -                          -                                        ^                     ^

_ _ w. 6. ana"v bu run

                                                                                                                                                                      ,m.                           _, m                      - - . ,__>               .m.
                                                                               --ePeRRE-s_,. .. s.....,,,....,..v.                    ,_
                                                                                                                                                                                                               .. nw . .u ni. 6 . ,. .. --

fu fizu a ^iu. p4: Gen-/

              ..                                 ,.                               MNote 3 modifies this SR by stating that                                                                                                                          s not required tb' '

O be set-the Required Acti

                                                                                                                             . wit tggvalve clos                                                                                            n accordance with
                               /                        8h*                                                                                             . . .                    ,
                                                                                                                                                                                   ,.,                                                                                                     O S.i'w
                                                                                                                                                                                                                                                                                           & 3'4.//                                                                                                                                                                "

5e jfoc

n; Valvejn' Q2.$1l-f

                                                                                                                                                                                                  >           ,r c                   o /Gn inco ester Ne SR 3.4.11.2                                                                                                         1-T.rk                     an Mi.rolable leak $w,
                                                                                                                                                                                                      -If,e AcS.cinae +Ae foRy(r alreJy SR 3.4.11.2 requires a complete cyc "N-                                                                                                                                    Operating a PORV through one complete cycle ensures that the PORV can be manually actuated for mitigation of an SGTR. M.: F :q=xy of h/c-3.f. coy
                                                                                  = _--eh:                                                               t i k=' ref                                                         cyci' er.d ir.dustry
                                                                                            .n<             h beeed cr.

a . .a < . - oe :d,y eyen*uelir.;;Aar e ne e s4swn +h.+ +here Valve.c u.run // y4a.rr ,c,-y-fle~Kiryghs;eiMen(),[*rfuemed =+ +he r*yfred Zn n+4 .. .. . ._ ........:.: , . _ , . . . . . . .

e .. , J a. . >,. ' . . . p

                                                                                                                                                                                                 .{.         . ..                .n..           ...Nn.       , . . . .    .II.
                                                                                    .J     +Sh'>      ..              .]'   .$    .

4 .$'.l g

                                                                                                                                                                      -1    8        , hh                    #                     ' . .. R'.-Og            4E.IP       ,      P   WIl
                                                                                    'l) 4 r'        .
                                                                                                         .....e_              _ . h t g y n s< s. . t . .     't..           ,,   e' ti sl p e k w_            . . s _ n, . .l   ., A g b v       _
                                                                                                                                                                                                                                                       .c..es          d e . ,(

4 a (continued) MARX UP 0F NUREG 1431 BASES B 3.4 61 5/15/97

                                                                                                                                                                                         - . _ ,                                    , , . . - . -                                 -,e.

CHANGE NUMBER JUSTIFICATION . 3.4 42 Not applicable to Callaway. See Conversion Comparison Table (Enclosure 6B).

3.4 43 Not applicable to Callaway. See Conversion Comparison Table (Enclosure 68).

            .4 44       Steam generator levels for MODES 3, 4, and 5 are specified to ensure SG Art             4     tubes are covered. The C;1h=; current TS did not ensure tube            f)7.f.5,~1 N *I'Y M 2 , overage.                                                  hfo}, udee 3.4 45   BITS 3.4.12 has been revised to move the No:e for Required Action B.1 regarding CCP pump swap operatio       and theFApplicability-et fer      g S, f:/2 . 2 g /Ty accumulator isolation to the LC0 ,M dir~r d " trr;;lcr "^^ 51-- Q 7.f g "r; 1.glant specific time allowances for exceeding the LCO's number
 %3* Nel*e dad          of [ECCS] pumps capable of injecting into the RCS are incorporated [ as l** NM2b               discussed in CN 3.4 18]. * = umter det:j, 31+ngt e = gyg73

=:;tMnr te th^ LC^ 270 p^ mitt;d med m c ,m,r; Oppr pri:t 'y ene bM A =n:t:ted und:r the LCO.

LCD t*f?*i &s_j)Consistent with current TS 3/4.1.1.4, " Minimum Temperature for Aff,, 3.4 46 Criticality," ITS LCO 3.4.2 and its Condition A and SR 3.4.2.1 are modified to refer to " operating" RCS loops. Adopting the current TS wording is acceptable since valid T,,, measurements are not obtainable for a non operating loop. 3.4 47 ISTS SR 3.4.11.1 contains a Note which exempts the cycling of the block valve when it is closed in accordance with Required Actions of Conditions B or E of LC0 3.4.11. However, Required Action A.1 also directs closure of the block valve when one or more PORVs are inoperable and capable of being manually cycled. The SR Note should also exempt performance when the block valve is closed in accordance with Required Action A.1 as the block valve should not be opened when the PORV is inoperable. This change is consistent with NUREG 1430 and NUREG 1432 inasmuch as the block valve cycling is exempted under Conditions A, B, and E. -Sk e per to the bh;k ;;h:3) t S E f.//--f -

                         =ht:ted t ". quired Actie.. n.1. Tfie Note to SR 3.4.11.1 will be 4 g gag                 revised to not require the surveillance performance if the block d

g valve (s) is closed per C=dithr ^ Sine p;wcr to th; block ch:3) 4 t rc=v;d in "equired Action; ".2 mid C.3, tre arvcilhne := not bc gf g

C han +% erding ch=g^ fr= "xt" t "perf ed" " the 2tc.-

the .;;rding of 5" 3.4.11.1.5 ensed to x;csed;t; the Conditi= B end E av--ti= This change is consistent with traveler WOG 87. b.rNxnr M-7A A note is added to ITS 3.4.8 ACTIONS indicating that entry into MODE 5 3.4 48 Loops Not Filled from H0DE 5 Loops Filled is not permitted while LC0 3.4.8 is not met. The addition of this note is based on the performance of a plant specific LCO 3.0.4 matrix (see CN 1-02-LS 1 of the CTS 3/4.0 package). %A/KRT 6A-78 4 f,f,#-/ JUSTIFICATION FOR DIFFERENCES TS 7 5/15/97

INSERT 6A-7A Q 3.4.11-4 In addition, Notes are added to Conditions C and F stating that these Required Actions don't I apply when the block valve (s) is inoperable solely as a result of its power being removed per l Required Actions B.2 or E.3 as a result of an inoperable PORV(s). In this scenario Condition B ' or E is entered as a result of an inoperable PORV(s). If one PORV were inoperable and incapable of being manually cycled, Condition B would be entered at time zero, to. Required Actions B.1 and B.2 would close the associated block valve and remove its power within time to + 1 hour. If, as a result of block valve power removal per Required Action B.2, Condition C were then entered, Required Action C.1 would require the associated PORV to be placed in manual control within time to + 2 hours. However, the reason for originally entering Condition B is that the associated PORV is inoperable and can't be manually cycled, thus there is nothing to be gained by placing the PORV in manual control. The PORV inoperability may be such that the PORV can't be placed in manual control (e.g., blown control power fuse), in which case Required Actions C.1 and C.2 can't be met. In addition, Required Action C.2 (restore block valve to OPERABLE status) can't be satisfied as long as power is removed from the block valve. Restoring the PORV to OPERABLE status within time to + 72 hours allows the plant to exit Condition B. If power were not restored to the block valve at this time, the new Note on Condition C would have no standing and Condition C would be entered. Similar conclusions can be drawn for the relationship between Conditions E and F. If Condition E is the original Condition entered, there is nothing to be gained by Required Action F.1 and Required Action F.2 can't be satisfied with block valve power removed. With F.2 not satisfied, Required Action G.2 would require the plant to be in MODE 3, but Required Action E.4 would have already had the plant in MODE 3 two hours earlier. I l l l l

ll $+.=. n -; CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431 Page 6 of 7 44 ,4 SECTION 3.4

                          -/+mvs/sr 7.Prr .2p4 DIFFERENCE FROM NUREG-1431                                                          APPLICABILITY                          ,

DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY , NUMBER [ DESCRIPTION

                                 /                           c%                                                                                                         '

3.4 45 L >ITS 3.4.12 has been rei ised to move the Note for Required Yes No - Operation of 2 Yes Yes Action B.1 regarding C(P punp swap operations and the Ng/e Q,,. CCPs is allowed per Applicability N::: L." accumulator isolation to the LCO,p, CTS. Also see CN __,__,

                                    - d! rr rd '- t = ' .. "~' 51 ^ _ . 1. dant-specific                         3.4-52.                            n   a.7-gv       -

time allowances for exceeding the LCO's r of [ECCS] $ f,6 /~2.--[. pumps capable of injecting into the R are incorporated [. > as discussed in CN 3.4-18). 3.4-46 ITS LCO 3.4.2 and its Condition A and SR l.1are Yes Yes Yes Yes modified to refer to " operating" RCS loops. 3.4-47 ISTS SR 3.4.11.1 contains a Note which exempts the cycling Yes Yes Yes Yes of the block valve when it is closed in accordance with Required Actions of Conditions B or E of LCO 3.4.11. However. Required Action A.1 also directs closure of the block valve when one or more PORVs are inoperable and capable of being manually cycled. The SR Note should also exenpt performance when the block valve is closed in accordance with Required Action A.1 as the block valve should not be opened when the PCRV is inoperable.$ 9//A%, gg y, gg./, CM/M r O d[:~~-/p Am -/- //4 3.4-48 A note is added to ITS 3.4.8 ACTIONS indicating tha$ y Nk &# 'f# #" es**W# /'dNVdfJ f$/#" C # W/' N 8'ie[ #'

Into H0DE 5 Loops Not Filled from H0DE 5 Loops Filled is not permitted while LCO 3.4.8 is not met.

3.4 49 This change reorganizes the presentation of ITS Yes Yes Yes Yes LCO 3.4.12. adds the word " required" to ITS SR 3.4.12.5. and changes the word " met" to " performed" in ITS SR 3.4.12.8. , , , 3.4-50 This change is consistent with current TS SR 4.4.9.3.3. No - Adopting ISTS No - Adopting ISTS Yes Yes The 12 hour frequency applies to vent pathway (s) that are format. format. not locked, sealed. or otherwise secured in the open position. The wording added to ITS SR 3.4.12.5 is also consistent with the format used in similar ITS 3.6 SRs. The 31 day frequency is also revised to be consistent with current TS SR 4.4.9.3.3. CONVERSION COMPARISON TABLE - NUREG-1431 5/15/97 -

l l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.11-6 APPLICABILITY; CA, CP, WC REQUEST: Difference 3.4-49 (Wolf Creek, Comanche Peak and Callaway) Comment: This difference does not address the addition of the "Immediately"in Required Actions D.1, E.1, and G.1 of ITS 3.4.11 FLOG RESPONSE: This RAI refers to JFD 3.4-49; it should refer to JFD 3.4-39. Difference l 3.4-39 is revised with the addition of the following: l 1

         "New initial Required Actions are added to Conditions D, E and G to immediately initiate actions for restoration of the inoperable PORV(s) (and/or PORV block valves) to OPERABLE status. These immediate actions will ensure expedient measures are taken to re-establish the operability of the PORV(s) (and PORV block valves) while maintaining plant conditions above MODE 4 but less than 500 F."

ATTACHED PAGES: Attachment 10, CTS 3/4.4 - ITS 3.4 Enclosure 6A, page 6 l l l l l l

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

l CHANGE NUPBER JUSTIFICATION but would limit the exception to prior to entering MODE 2. This change is consistent with traveler CCO. - TJrF-air. d) f.f //-a 3.4 36 SR 3.4.13.1 and ACTIONS for LCO 3.4.15 are revised with the addition of - a note per traveler TSTF 116,Pr;. 1. The note addresses the concern f) f.f,/f-2 ! that an RCS water inventory balance cannot be meaningfully performed unless the unit is operating at or near steady state conditions. The Note added to the surveillance provides an exception for operation at less than steady state conditions. The RCS water inventory balance will only be allowed to be deferred for 12 hours after re establishing steady state conditions. 3.4 37 The primary to secondary leakage limits are revised per Callaway OL Amendment No.116 dated October 1,19%. 3.4 38 tent with traveler im 105, the details on - nuc Dy which 1 the RCS is verified are moved f 3.4.1.4 to the Bases. Moving this informatlo ows the use of precision heat balances, elbow taps, r acc thods in order to perform this verificat nd is consistent with the - ev. 1 philo moving clarifying information and descriptive s out w Ts to the Basess y fy /, f d, , f, , ' f) 7.f,/- / Onvention Enc /s.run $8 3.4 39 The shutdown %nrironTible reqdirements of ITS .4.11 would requi the plant to reduce T., to < 500*F within 12 hours, rather than go to MODE 4. to address the concern of entering [COMS] LC0 3.4.12 Applicability with inoperable PORVs. For consistency, the shutdown requirements of ITS 3.4.16 are a o revised to allow 12 hours to reduce L. to < 500*F. This change is e

sistent with traveler TSTF 113/.~.fA/JEgP

                                                              ^ -                  ^
                                                                                              "                 ~

gA- f] 8 M //~S 3.4 40 Consistent with traveler 4dO9-99 1 77-.-;272'o Vthe Note t SR 3 ... 414 would be 8 S.4- . / .2 l modified to provide additional time to perform an RCS precision flow rate measurement. The time allowed would be changed from 24 hours to l 7 days. This change is acceptable because other indication of RCS flow is available (SR 3.4.1.3. RCS total flow meters) and additional time normally would be required to establish plant conditions suitable for the precision heat balance. Since this parameter does not normally change significantly and the flow meters can be used in the interim. there'is no need to perform this SR within the 24 hour period specified in NUREG 1431 Rev. 1. The 7 day period provides sufficient time to establish steady state plant thermohydraulic conditions and obtain equilibrium xenon. In addition, the THERMAL POWER specified in the Note would be changed from the generic value in brackets (90% RTP) to 95% RTP, This change is acceptable because it specifies a power level in better agreement with current operating procedures for performing a precision heat balance. Current TS do not specify a power level for this measurement. l 3.4-41 Not applicable to Callaway. See Conversion Comparison Table ! (Enclosure 68). JUSTIFICATION FOR DIFFERENCES TS 6 5/15/97

_ . . _ .., _ . _.-.. . -._ .~._._ ~ _ . . .._.. ._ .. ...._ .- .. . ...... ~-._. . . . . . _ . . . INSERT 6A-6 . Q 3.4.11-6 New initial Required Actions are added to Conditions D, E and G to immediately initiate actions for restoration of the inoperable PORV(s) (and/or PORV block valves) to OPERABLE status. These immediate actions will ensure expedient measures are taken to re-establish the

                 - operability of.the PORV(s) (cad PORV block valves) while maintaining plant conditions above                                   ;

MODE 4 but less than 500 F. I 1 l l t f i 4

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.12-1 APPLICABILITY: CA, CP, DC, WC REQUEST: Difference 3.4-49 Comment: WOG-100 has not yet become a TSTF. l l FLOG RESPONSE: WOG-100 has been approved by the TSTF and is designated as TSTF- l 280. This traveler has been submitted to the NRC and is under review. The proposed wording l in TSTF-280 was modified from WOG-100 and these modifications have been incorporated into the ITS (added "or" to LCO list and SR 3.4.12.5 Note was deleted). The FLOG continues to pursue the changes proposed by this traveler. i ATTACHED PAGES: l 0, CTS 3/4.4 - ITS 3.4 ) Enclosure SA, Traveler Status Sheet and pege 3.4-28 Enclosure SB, page B 3.4-78 A, page 8 ! l I l

TRAVELER # STATUS DIFFERENCE # CO M NTS CO Incorporated 3.4 40 Applicable to Callaway and Wolf 1 737/~,ar:, _ _ creek only.42f./ . . . , _ , , [ 4e@$dif-"# Incorporated 3.4 49 [ 7.1:/.2-/ f M 7.f ;;7 W. __ _ l i l l 1 l j s MARK UP OF WOG STS REV 1 (NUREG 1431) 5/15/97

                                                                                                                                                   .u           .: ps
 ,                                                                                                                                 w a vi     ry -e wwm i

4 <_.- 3.4.12 ' e SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY m Verify RHR suction aMR, valvel +s E open E 1-2 ur 3.4 08 for each required RHR suction relief valve. ,- Q 3'. O-3 -- -- ~

                                                  ~                                                                                                 /-

SR 3.4.12.5 = -- -- -

                                                     -.---:ru'E .--- .............--                                  - - -

f) d./3-/ a o 1 ,, ,,

                        'My mted te M pet: ..d fr. cr;';t;                                                                                              < ~/

uth '_C0 3.' .12 2 ? 3.3.4-49 l N _ - .................. - V - -

                                                                                                                                                             -3;4 49 Verify M RCS vent : -                                    0-07 square                        12 hours for                         . B.ps e
   .                         inches open.                                                                                unlecked ega                      .c3.4 50 vent           ' .

velvc(s) anammmmmmmman M

                                                                                                                           ...a       ,     . ..
                                                                                                                          , ,n , , , u .. i .

ale 31 days for locked aga 3.4 50 s vent valve (s) i . s , .e. .,. n . :. IBEMinREERI 3 ;i, , ~. m ... , .7 SR 3.4.12.6 Verify PORV block valve is open for each 72 hours required PORV. SR 3.4.12.7 EllEEEl '.'crify essocietcd "Jn ;uctier 31 days 2. t 00 W isoistion velv; is 1;ckcd ega with egreter g jog ( r ,j/,%,) gucr re.,.vcd for cech rcquired "JR suction

                             ..,u      ..      ..                                                                                                       f) 3.+, /2-3 Ib5the     TU1Th.

(continued) MARK UP OF WOG STS REV 1 (NUREG 1431) 3.4 28 5/15/97

i, he I %FI e#J e WI51 4 B 3.4.12 BASES I c n ~ > . ~ .. SURVEILLANCE (continued)

REQUIREMEKTS isolation valves are open and by testing it in accordance with the Inservice Testing Program. (";fer t; 2 3.4.12.7 fer tra "J: ",

satim is;1 stim v;1ve irnillerce.) This Surveillance is only required to be performed if the RHR suction relief valve is being used to meet this LCO. The R}R suction m valveg 4s Eg verified to be opened every te M hours. The Frequency is acnsidered adequate in view i of other administrative controls such as valve status indications available to the operator in the control room that verify the R}R suction . -r - valveg remaint open. The ASME Code, Section XI (Ref. 8), test per Inservice Testing Program verifies OPERABILITY by proving proper relief valve mechanical motion and by measuring and, if required, adjusting the lift setpoint. a SR 3.4.12.5 The RCS vent of x ' e:47 square inches is proven OPERABLE by verifying its open condition either: l

a. Once every 12 hours for a velve - ~ that eenrue be - locked - .- . * . -

a 11 ;i5 t . , -

b. Once every 31 days for a valve that is locked, sealed, or 1 .. 3- secured in O 7 ' position. A removed pressurizer safety valve J- I nd ' i . # - fits this category.

The & passive vent m- arrangement must only be ope gg 6.?,f/2-/

                                              """8            o be OPERABLE. This Surveillance is required                                                      py,.l q--

j@,j,y. f the vent is being used to satisfy the pressure relief r uirements of the LCO 3.4.12kg (continued) MARK UP OF NUREG 1431 BASES B 3.4 78 5/15/97

CHANGE I NUMBER JUSTIFICATION 3.4 49 LC0 3.4.12. "[COMS]" provides four different methods for pressure relief. Any of the four methods may be used. However Surveillance l Requirement 3.4.12.5 requires testirig whether or not the equipment is being credited to meet the LCO. The proposed change adds the word l

                    " required" to the Surveillance to exempt its performance if the equipment to be tested is not being used to meet the LCO. In addition, l                    two editorial changes were made. The LC0 requirement presentation was              j i

clarified. Also, the Note to SR 3.4.12.8 was revised to replace i " required to be met" with " required to be performed" since the L " performed" nomenclature is appropriate here consistent with the CTS. This change is consistent with traveler ET 100. / a7F-:::r/0, $7A/J-/ 3.4 50 This change is consistent with current TS SR 4.4.9.3.3. The 12 hour , frequency applies to vent pathway (s) that are not locked, sealed, or I otherwise secured in the open position. The wording adde,i to ITS SR 3.4.12.5 is also consistent with the format used in similar ITS 3.6 SRs. The 31 day frequency is also revised to be consistent with current TS SR 4.4.9.3.3. 3.4 51 Not applicable to Callaway. See Conversion Comparison Table , (Enclosure 68). 3.4-52' Not applicable to Callaway. See Conversion Comparison Table (Enclosure 68).

g. +.ss .1 xceer 2.+-ss lac-3447

S?+-s+ WJ yahDa de Ody. See Csnvenron (%p,,,,, bc-3.4

                     %Ala,[encle.ruray).

L JUSTIFICATION FOR DIFFERENCES TS 8 5/15/97 l l

1 ADDITIONAL INFORMATION COVER SHEET i ADDITIONAL INFORMATION NO: O 3.4.12-2 APPLICABILITY: CA, CP, DC, WC , REQUEST: Differences 3.4-20 and 3.4-45 l Comment WOG-51 Rev.1 has not yet become a TSTF. i l FLOG RESPONSE: WOG-51, Rev. 2 is designated as TSTF-285. This traveler has been l submitted to the NRC and is under review. The proposed wording in TSTF-285 was modified ! from WOG-51, Rev. 2, and these modifications have been incorporated into the ITS. The j FLOG continues to pursue the changes proposed by this traveler. 1 ATTACHED PAGES: Attachment 10, CTS 3/4.4 - ITS 3.4 Enclosure 2, page 3/4 4-34 Insert B Enclosure 5A, Traveler Status Sheet and pages 3.4-24 and 3.4-25 Enclosure SB, pages B 3.4-71 and B 3.4-72 Enclosure 6A, page 7 l Enclosure 68, pages 6 and 7 l i 1 l' l

_. _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ . . _ . _ .~.- _ _ _... l , INSERT A FOR PAGE 3/4 4 34

I with a maximum of zero safety injection pumps and one centrifugal charging 9-et-Af l pump capable ofinjecting into the RCS and the accumulators isolated or f-/s.-pf L depressurized below the allowed RCS pressure per the PTLR INSERT B FOR PAGE 3/4 4-34 7 x

                                                                --NOTES--- g--------------                                                        7-4_J

1. Two centrifugal charging pumps may b# capable ofinjecting inte +e RCS & 7.f./n-2 for 5 4 hours for pump swap operations.

1 l

       \    en4/e capable ofinjecting;inic ic RCS-2. Two safety injection pump (a)In MODE 3 with any RCS cold leg temperature 5 368 F and ECCS pumps OPERABLE pursuant to LCO 3.5.2, "ECCS-Operating",

and i I (b)For up to 4 hours after entering MODE 4 from MODE 3 or until the temperature of one or more RCS cold legs decreases below 325 F, whichever comes first.

3. One or more safety injection pumps mayVbe capable ofinjecting Mc ic 4.zf./m-2 ;

                  -RE&in MODES 5 and 6 when the RCS water level is below the top of the

!. reactor vessel flange for the purpose of protecting the decay heat removal funedon. j, ag,,j,pj

4. Accumulator inchtica n; caly rcqui :,,8when accumulator pressure is gj,j,,_3 ger::ir.n or equr.! : the maximum RCS pressure for the edsting RCS l l cold leg temperattr e a owed by the P/T limit curves provided in the l PTLR.

                                                           /an %                                                                               ,
                                           /                                                         _

t i , IMXISTRY TRAVELERS' APPLICABLE TO SECTION 3.4 . l . TRAVELER # STATUS DIEEERENCE # COMENTS - TSTF 26 Incorporated 3.4 32 Approved by MC. 4 TSTF 27, Rev. [ Incorporated 3.4 33 Ajpener/lyA/4d -((k j, I' TSTF 28 Incorporated 3.4 22 Approved by NRC. TSTF 54, Rev. 1 Incorporated NA d$s"Ob37cdhnly. ry-y4 49 x ! T5TF 60 Incorporated 3.4 15 Approved by EC. y' TSTF 61 Not Incorporated NA Minor change that is adequately I addressed in the Bases. R 1 TSTF 87, Rev. ncorporated 3.4 31 A//rsvMyl NAO - 7p 34y ( , TSTF93,Rev.$ Incorporated 3.4 17 M $r"o M /or N a'way in OL i ! L3 4 2+,9-3 Amendment No. 105. g af ; i TSTF 94;gev, l Not Incorporated NA Retained current 13. 7p_y,4f k k

             = 1~ J=. !           : =. ,~, =                  3.a                            g241 i ssn:i                    :

TSTF 108, Rev.1 Not Incorporated NA LCO 3.4.19 does not apply. TSTF113.Rev.[Yncorporated 3.4 39 8 7.4tf-/[$2f,//-3 3 7,' 3/j - f i TSTF 114 Incorporated NA Approved by RC: Bases 3.4.7 changes only. 4 13TF 116. Rev. [Yncorporated 3.4-36 j9;f.4/7-2 3;,y gll l TSTF 136 Incorporated NA B E 35 U [df U .I changes on1y. y g-y4eq\ BM N Nch7h nly. q_g4 g9 TSTF 137 Incorporated NA i TSTF 138 Not Incorporated NA Inconsistent with RCS loops j requirements of ITS 3.4.5 and 3.4.6. l TSTF 151gev. / Incorporated NA Bases 3.4.11 changes only.7hS.ANk TSTF-153 Incorporated 3.4 01 [/Pr*VM k NAC- 74-%f#1 _1WTSTF Incorporated NA EsM.kchIn$>'only. 72-3&g, l [ ggm-air Incorporated .2 M C 3.4 45, See also bsi'.T'-[a#nd3N6.[

Uk '

             --                    Incorporated                3.4 35                     g y,4g-p g 7.f ;jp h b% 8 DsN-                          Incorporated                3.4 10         DCPP only.$r,wllyN#d,gs.q@
;            WOG87,/ev.:2.         Incorporated                3.4 47                     gIf://-f      73" 2.f mg

{ MARK UP OF WOG STS REV 1 (NOREG 1431) 5/15/97 [t-

ps .

                                                                                                                                                                                 , ,,m
   ,-                                                                                                                                                                             n. i m      w, .,m wwm 3.4.12 3.4 REACTOR COOLANT SYSTEM (RCS)                                                                                                                                                                                      .

l 3.4.12 --:.~-. . - . . . . . . ....t . . . e La T; 4 r;tur; pS' Overpr;;;ur; "retectier. (LT0"' S,;t;;;; 1 LC0 3.4.12 ,. . m . m , , . . .- shall be OPERABLE with a maximum of M , PS o m... . . . . . and . ;~. .- --w. .. < capable of 2-B PS r injecting into the RCS and the accumulators isolated and ;itt.;r ;

                                   ;7 t tel;w.                                                                                                                                                                 !3.4 49--
                                   ;.            Tw; "CS reli;f velv;;, ;; f;11 a;.

t- 3 Two power operated relief valves (PORVs) with lift settings within the limits specified in the PTLR, or e- E Two residual heat removal (RHR) suction relief valves with setpoints 2 v. psig and s M psig, or : B; ::i S- One PORY with a lift setting within the limits specified in the PTLR and one Rm suction relief valve with a setpoint ,,,, a . - psig and s . . psig g 4Bf b- 3 The RCS depressurized and an RCS vent of a M E-97 square - B PS2 inches.

                                                                                                                                                                                                               ;3.4 45s Md*                                                                     , , , ,                                      !

3 '

                                                                                                                                                                                                              .:3.4 18                l t ' ' ' ' N 3.4./2- 2                                j

c. . . .. . . . . , .. .. . .- . . . . . . . . ... - , . . . . . . . , .. . . - - --3.4 18x

                          &e-                                                                                                                                                                            Q'?.4J2-2 A

g 4 . . .. ,< .. .. .. . . . - .., . . . ..

                                                   . . . < . w: o . a . ..c.g.-                        .. ..            ..a..           .a. ..                  .,1.....             ....            .

g ,.c ... ~ , . . . . .

                                                                                            ...f       . .n . .c . , , . , , . . ,7
                                                                                                                                        - .: . ,              .t;.s.m.                     . . .

v.x...,...;:.

                                                                                             ; , _. . . . . a :.y..               .

a. , .- . n . :, . _

                                                                                                                                                                                                  .m 4f.(S.a.          . .- ' .; - e A d ,. 3 ..
                                                                                                                           / bade g       ..      ...       ..-       : , . . .     ..a.         .g. y . 3 9            ...    .. : .     .i,y.g.           ..,          ,,..c..4              .

3.4 20'

                                                                                                                                                                  .o,c
                                                        , .      3_:      . o..      .

5,.,. n... :... , i ,y , m .i . . . . jg

                                        ..     . 9, . . . . u        o . y ,..                   cn   .(.      . . a . .tg .,, 4, . .;,,y .;                    n .        ... ..         .s t
                                      ~
                                               ' , ,'\;I j i ' 8 Iw ' . f *I. r s...t-(continued)

MARK UP OF WOG STS REV 1 (NUREG 1431) 3.4 24 5/15/97

l l i j t PS: j -.w, y . b- ; i , my Se unia/a/e) 3.4.12 l 1 I E -.3.4 45e j ,4 f.f: /2-2 ? ~ L/ err + fan i 1 1 1 APPLICABILITY: - -. o a-;-.;. s: - o n., -

                                                                                                                                             . .. .                                             +3.4 18e
!                                                                    HODE 4 0.;n eli "00 ;;1d 1;; te rcretur; i; a 275'",

! MODE 5. { MODE 6 when the reactor vessel head is on.

,unw-

                                                                                                                                         ~.m l                                                                     A;;;;leter i;;1; tion i er.ly r; quired J.;r. ;;;; uleter                                                                  e3.4 45'

! pre;;;r; i ;7;.t;r it.; r ; ;-l t; tt.; si;;;; "CS pr;;;;.; for th; cxi;tir.; "00 ;;1d 1;; t; yer;tur; ell;;ed by th; __mm, l',"

                                                                     ,2_2m    _ . _ . . _ _                         2_     me-         o,, .

5 .u.b bwl ybe y1 w T .www 555 b..b 5 . b.n . 1 x ! ACTIONS 1 1 [t - , G) p

                                           ..           sp . ..    .. ,

c..- . 3;4-30 -- l ( .t 4 . l CONDITION REQUIRED ACTION COMPLETION TIME 1 i 1 l A. Two M or more - G A.1 Initiate action to verify a Immediately i , se- pumps capable maximum of ~ . o . :3.4 06.

!                                               of injecting into the                                      m' -- ~ . pumpg+s .
                                                                                                                                                                                                   , g.pS 3 l                                                RCS.                                                   capable of injecting into the RCS.

t 1 l E Two ee-sere < - " . u. . - B.1 "T

                                                                                                                                       ~                                                              B.PS l

charging pumps capable Tw; cherging p;;p; my k ;.3,4 06 of injecting into the ;;p;bic ;f inj;; ting int; 3.4 45-l' RCS. th; "0S during pu;p ;ng ep;retien fe a 15 sir.ut;;. f 1 i Initiate action to verify a Imediately j maximum of M B PS= r charging pump is capable of

injecting into the RCS.

4 1 l; (continued) i l MARK UP 0F WOG STS REV 1 (NUREG 1431) 3.4 25 5/15/97 t 4 1 i

4 a .. l L T Sy;t;; B 3.4.12 ) 1 1

BASES APPLICABLE RCS Vent Performance SAFETY ANALYSES (continued) With the RCS depressurized, analyses show a vent size of ,a E: 07 square inches is capable of mitigating the m a";nd LT 9 ;.;ri,ressure transient. The capacity of a vent this size is

. greater than the flow of the limiting transient for the L-TOP M i configuration, era llP ps;p er.d . ' A e mt .-,+ o? 1 OPERABLE, maintaining RCS pressure less than the maximum pressure on the P/T limit curve. 4 The RCS vent size will be re evaluated for compliance each time the P/T limit curves are revised based on the results of the vessel material surveillance. The RCS vent is passive and is not subject to active failure. i The LT Sy;ts; M satisfies Criterion 2 of tt.e = P;1 icy eu LCO This LCO requires that the L T Sy;te; ~ is OPERABLE. The LT Syste; M is OPERABLE when the n+nfetse w coolant input and o.: ? pressure relief capabilities are OPERABLE. Violation of this LC0 could lead to the loss of low temperature

overpressure mitigation and violation of the Reference 1 limits as a result of an operational transient.
To limit the coolant input capability, the LCO requires -

7, r- :. .n c. <. . . , . v ., . ;7,; lr: p;;p , . . . 4,, . e . n. . .. . . , u . ., . . .. .? capable of injecting into the RCS and all accumulator ischarge isolation valves 3 closed and ggjg'"j

                          -+-           Tumobilizedl                                 ,

accumulator pressure is greater than of equal to the maxi RCS pressure for the existing RCS cold leg temperature allowed in the PTLR.^ s made !

                                                                                                     -}-ow C 3'4 -M a, Q 3.+,/2 i                                                c.j,:..v   .
                                                                      ,; ,,:., m .a. 2- wa rrc,-                     a .: - . c.-.;<,e         m
                                               ;_w-            pey.;                   .9   eau.g . 9 , ; . i- . - . n . .x. c:,:
                                                                                                                                           . M.o j                                               thatBUIRBpp,u; :: a r a sin d n6MO;-W 9 M J

i

!                                                                                                                                      (continued)

MARK UP OF NUREG 1431 BASES B 3.4 71 5/15/97 f

i 1 ]

                                                                                                                                                                                                               , ,n.o

5. s 5 e.._i_

                                                                                                                                                                                                                              =#,7 e wwass B 3.4.12 ntinued)                                                                                                                                                                                                                        I j                                                                                                              '                  f 4                                                   .i4          '.        ...,s.4            _g...               i.,s..,<..                     .., . . . . .. . .; . . ' . ,                                  . . . . <           .
                                                 .i.., , : -                                                                                                                  i.

.e ir i t. . .i . ' : . . . . 4 y ., , . . . . .y .# ,.

Q, ' ' N A N2-2 g ..

                                                                    .        40.
                                                                                         ~
                                                                                              - ., t si t!      i;., ? *~:               .......I.... ..,'..e
                                                                                                                                                                                      .d.       . - ..'.:                    .p, 21 p J[ i - - 4, ' Kf @ .~ ' - . i ' r ").: I""
                                                                                                     .                                                 ** -       P.'.*1'-
                                                                                                                                                                       /

I * -' b ,.:.?..u.;f. ;, ; 1 i i 1 . . ,, s. , ,. - . . . ..,,. . . . e! - - . . j , .. .. ... p ~, .- , . , , .

                                                                                                                                                                  .               r.            . F    ':
                                                                               . i ,< .. ,:,                                              ., .1 i
                                                                                                                                 . . . . ~ , ..                        . . . . ,.            .       ,

g , .T 1 , i . a '} ' - . ' ,

                                                                                                                                                             .+.. :                  4 -
                                                                                                                                                                                                                      ,....,..                ((/Q 2
                                                 ;8 .               -.              . , ,         n: .             . . . . . . ..

._ .~. . . . . .

l -Me **f A s anisolaled

. . . . . . .. ... .. .. u. ,i. .... .

3 Mpeup puluulmuuuukv " "'" ^ i m ll L.lur & 4 The elements of the LCO that provide low temperature overpressure j mitigation through pressure relief are: i i

                                                                   , _        m,,           . _ _ , , _ ,                                _ _ , _ , , _ . . .
                                                 ..                 .,-.s...                                 . , ......

l l 51- Two OPERABLE PORVs: or ! A PORV is OPERABLE for t-TOP Ss when its block valve is

open its lift setpoint is set to the limit required by I the,PTLR and testing proves its ability to open at this I ,

setpoint, and motive power is available to the two valves i and their control circuits. I i I (continued) ! MARK UP OF NUREG 1431 BASES B 3.4 72 5/15/97 t

CHANGE NUMBER JUSTIFICATION 3.4 42 Not applicable to Callaway. See Conversion Comparison Table (Enclosure 6B). 3.4 43 Not applicable to Callaway. See Conversion Comparison Table (EnclosJre 6B).

            .4-44                Steam generator levels for MODES 3. 4. and 5 are specified to ensure SG bn             *'

tubes are covered. The C:11=; current TS did not ensure tube f) 7. f. 5 ,"1 N *I"Y M 2 , overage. hf,.l, u,,}ee 3.4 45 BITS 3.4.12 has been revised to move the No:e for Required Action B.1 regarding CCP pump swap operatio and theFApplicability St for - g S, f;/2-,:2 rega /T7 accumulator isolation to the LCO .M disc"-:M " tr=1;r C 51 - Q 7, f ;g "c; 1.p/lant specific time allowances for exceeding the LCO's number

    %#8 Alel*r Add               of [ECCS] pumps capable of injecting into the RCS are incorporated [, as l8'M N er/ M r             l discussed in CN 3.4 18]. ^ ^ - eter Stri' siM'?t h = .;hcre
     #I          /               :=:pther t tP: LC^ cre        p mitt d mra m o a r; :ppr:prht:ly A r=t:t:d =cr the LT.

Y*% r*v'Y 'j' q A/f,, 3.4 46 Consistent with current TS 3/4.1.1.4, " Minimum Temperature for Criticality," ITS LCO 3.4.2 and its Condition A and SR 3.4.2.1 are modified to refer to " operating" RCS loops. Adopting the current T'. wording is acceptable since valid Tm measurements are not obtainable for a non operating loop. 3.4-47 ISTS SR 3.4.11.1 contains a Note which exempts the cycling of the block valve when it is closed in accordance with Required Actions of Cond'tions B or E of LC0 3.4.11. However. Required Action A.1 also directs closure of the block valve when one or more PORVs are inoperable and capable of being manually cycled. The SR Note should also exempt performance when the block valve is closed in accordance with Required Action A.1 as the block valve should not be opened when the PORV is inoperable. This change is consistent with NUREG 1430 and NUREG 1432 inasmuch as the block valve cycling is exempted under Conditions A. B, and E. -Si= ~~ce- to the bb;k V:h:':) H $ ~f.f.//-+-

                                 =ht:id in ":;ir:d Actien A.1. "Ifie Note to SR 3.4.11.1 will be gag               revised to not require the surveillance performance if the block 4                           valve (s) is closed per'#C=dithr ^ Sin p=r to the bk;k V he':)

p g g g, S rc=v;d in "e cired Actions ".2 and E.O. tte arveilh=0 := =t b;

rd- tM erdhg chr;c fra "at" t; "p^rf =d" " the "0te.

the .;;rdt; Of S" 3.4.11.1 a is ned to ;;ces;det; the Caditi= S eM E av^-tix 4This change is consistent with traveler WOG 87. L .rNJWA'1~ M-7A 3.4 48 A note is added to ITS 3.4.8 ACTIONS indicating that entry into MODE 5 Loops Not Filled from MODE 5 Loops Filled is not permitted while LC0 3.4.8 is not met. The addition of this note is based on the 3 performance of a plant specific LCO 3.0.4 matrix (see CN 102 LS-1 of the CTS 3/4.0 package). .TA/MF 6A-78 4 f.ff-/ JUSTIFICATION FOR DIFFERENCES TS 7 5/15/97

n_ r r=v. CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431 Page 6 of 7 gg ,.4 SECTION 3.4

                      -/*.v,/,,- 7.cm:.at$

DIFFERENCE FROM NUREG-1431 APPLICABILITY NUMBER)DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY 1 c+ 3.4-45 b >ITS 3.4.12 has been re, 1 sed to move the Note for Required Yes No - Operation of 2 Yes Yes Action B.1 regarding C(P puup swap operations and the N ,/c Q,,. CCPs is allowed per Applicability N:t: ".." accumulator isolation to the LCO,p, C15. Also see CN

                                " di r ::d " t = :L = Si A . 1.                              dant-specific                        3.4-52.                               n      o.7_g y7.

time allowances for exceeding the LCO's punps capable of injecting into the R are incorporated [. r of [ECCS] $ f.A /2_.2 as discussed in CN 3.4-18]. 3.4 46 ITS LCO 3.4.2 and its Condition A and SR 4 .1 are Yes Yes Yes Yes modified to refer to " operating" RCS loops. 3.4-47 ISTS SR 3.4.11.1 contains a Note which exenpts the cycling Yes Yes Yes Yes of the block valve when it is closed in accordance with Required Actions of Conditions B or E of LCO 3.4.11. However Required Action A.1 also directs closure of the block valve when one or more PORVs are inoperable and capable of being manually cycled. The SR Note should also exespt performance when the block valve is closed in accordance with Required Action A.1 as the block valve should not be opened when the PORV is inoperable.$ g//AV,,,, g g g gg -f, Ojfg,, g gj f f,. , /. d.IM//-Y, 3.4-48 A note is added to ITS 3.4.8 ACTIONS indicating thaNNy h ke*s M l8s!#*~ O 'bie[ #

into MODE 5 Loops Not Filled from H0DE 5 Loops Filled is not permitted while LCO 3.4.8 is not met.

3.4-49 This change reorganizes the presentation of ITS Yes Yes Yes Yes LCO 3.4.12. adds the word " required" to ITS SR 3.4.12.5. and changes the word " met" to " performed" in ITS SR 3.4.12.8. ,, 3.4-50 This change is consistent with current TS SR 4.4.9.3.3. No - Adopting ISTS No - Adopting ISTS Yes Yes The 12 hour frequency applies to vent pathway (s) that are format. format. not locked, sealed. or otherwise secured in the open position. The wording added to ITS SR 3.4.12.5 is also , consistent with the format used in similar ITS 3.6 SRs. 1 The 31 day frequency is also revised to be consistent with current TS SR 4.4.9.3.3. , CONVERSION COMPARIS0N TABLE - NUREG-1431 5/15/97

         ,a                                                                                                                                                                b CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431                                                  Page 7 of 7 SECTION 3.4 DIFFERENCE FROM NUREG-1431                                                                        APPLICABILITY NUtBER      DESCRIPTION                                                               DIABLO CANYON         COMANCHE PEAK    WOLF CREEK            CALLAWAY r D4~ bc- ALL-MS The Note for SR 3.4.1.4 is remov M. This is consistent                    Yes                   No               No                     No 3.4 51 with DCPP CTS 4.2.3.5. DCPP con  ts a measured RCS total flow rate verification on ag       th frequency.                                                                                            ,,,

3.4-52 Consistent with traveler Z ".c.ENe#concerk No - See CN 3.4-45. Yes No - See CN 3.4-45. No bee b S - 5 accumulator isolation 1sg moved from the Applicability to the LCO. reade)Se clntI}y edif

f.4-S3 .ZNTER1~ 2. +-f3 leS No es ler WC-36-M f SA-s+ rwsexr 7.+-sy. y'a No Als A/s bc-z4.a,g r t k CONVERSION COMPARISON TABLE - NUREG-1431 5/15/97

1 ADDITIONAL INFORMATION COVER SHEET I I ADDITIONAL INFORMATION NO: Q 3.4.12-3 APPLICABILITY: CA, CP, DC, WC l REQUEST: = Difference 3.4-09 1 Comment The difference does not adequately justify not adopting STS SR 3.4.12.7. The SR is intended to apply to valves besides manual valves. Performing SR 3.4.12.4 does not verify the same status as that verified by SR 3.4.12.7. l L l FLOG RESPONSE: JFD 3.4-09 is not applicable to DCPP. JFD 3.4-09 provides an incorrect justification for not adopting STS SR 3.4.12.7. The Surveillance Requirement to verify the RHR suction isolation valves are locked open every 31 days (when the RHR relief valves are being used for cold overpressure protection) was removed from the CTS as part of a license amendment implementing the Generic Letter 88-17 l recommendation to delete the RHR autoclosure interlock (ACl). The 31 day surveillance was l- determined to be no longer necessary since removal of the ACI eliminates the single failure l that could have isolated both RHR suction relief valves. ACI removal also reduces the probability of closure of the RHR suction isolation valves when power is available. l Also, SR 3.4.12.7 is bracketed in the STS. NUMARC 93-03, ' Writer's Guide for the i Restructured Technical Specifications" indicates that brackets are used in the generic ! Technical Specifications and Bases to indicate where plant specific input is needed. As identified in the " Methodology for Markup of NUREG-1431 Specifications"in Enclosure 5A, I changes to bracketed information involve the insertion of plant specific information which is l presently located in the current TS. The methodology applied by the FLOG was that a JFD was not required if the bracketed requirement /information was not in the current TS. Therefore, no justification was provided since STS SR 3.4.12.7 was not in the current TS. ! SR 3.4.12.4 is also bracketed in the STS. The changes being made to that surveillance involve plant specific wording changes (i.e., " isolation valves are"), which require no justification per the FLOG methodology, and the SR Frequency as discussed in JFD 3.4-08 l (not applicable to DCPP). l Based on the above, Difference 3.4-09 is no longer necessary and will be replaced by "B-PS" in the Enclosure 5A markup. Difference 3.4-09 will be shown as "not used" in the Enclosure 6A and 6B markups. Plant Specific Discusaion ACI deletion and elimination of the subject surveillance requirement was approved for

              ' Callaway in OL Amendment No. 42 dated March 27,1989.

l i

f l i ATTACHED PAGES: Attachment 10, CTS 3/4.4 - ITS 3.4

                                                                                      . Enclosure 5A, page 3.4-28

! Enclosure 6A, page 2 Enclosure BB, page 1 l l i l l [

l I i

M j . 2 ps .

n. s vs wgawwm i 3.4.12 1

i i SURVEILLANCE REQUIREENTS (continued) i. 1 SURVEILLANCE FREQUENCY

             . .         m.                 Verify RE suction m valvel 4e i                                                           open            10         ur                    3;4 08.:

for each required Rm suction relief valve. ,.-

a Q 3. /2-3 -. ~ i SR 3.4.12.5

                                                                             ------ E                         ---- - - ---              -.   -
                                                                                                                                                                                   & 9.Y ./.2 - l 4

m ,,, 4 o u u ___,_ m .m -

                                                                                                                               - , u __                                        -

s,,so 2 -.,

                                                                                                                                                                                                .,~,,f 3
                                            --,..c                 ,."--r--                   ----""r                        ea eim*

e3.4 49 i ~ ~~ ~ ~ ~ ~~ ~ ~ ? .................... e3:4 49 Verify < RCS vent a 9-97 square 12 hours for : 8 PS,

l. inches open. enleek;d egn +3.4 50; i vent .

W . 8 7 5 \ eF f M l MM e *1 days for i leck;d egr. 23.4 502 i vent valve (s)

)                                                                                                                                               ..        ..        ..

3 i l. i SR 3.4.12.6 Verify PORV block valve is open for each 72 hours l required PORV. i { SR 3.4.12.7 M Verify ;;;;;iet;d %0 ;;;tica 31 d;y; 2.0 0^ j= W isci;tica ;;1;; i; led d egn witt, egr-ter g jog i ( ,-,j/ ,%, pa; rew;;d fer ced r;;uired %E ;uction __,2.._,. p 3.f,/.2-3 4 (continued) 7 MARK UP OF WOG STS REV 1 (NUREG 1431) 3.4 28 5/15/97 ,1 i

CHANGE MHlER JUSTIFICATION l 1 i 3.4 08 The existing licensing basis as contained in the Technical Specifications requires performance of this surveillance on a frequency of 72 hours. The Westinghouse STS Osed to develop the plant specific TS did not address the use of RHR relief valves. The requirement in the current TS was developed as part of an LAR to remove the autoclos'ure interlock which, in part. proposed 72 hours as it was l consistent with the SR for the pressurizer PORV block valves. The 72 hours was found to be acceptable in the SER which was enclosed in the license amendment. Plant experience has not indicated that the existing requirement is unsafe or unacceptable. The surveillance frequency does not require reduction to 12 hours. l l 3.4 09 % ;'irt dxs =t by =n=1 :"' ncti= i:chti:n =h=. 'M

                 =t:r ;per:ted n:ti= ichti= =h= (2 per r: lief nix line' cre
                  =r=ilkd in ==rt=: witt, " 2.i.12.t. T hr:f r tMs arxilkna mir x.^,t 2 =t = d. - /%+ u. red.                                     0 7.4.M .*E I

3.4 10 Not applicable to Callaway. See Conversion Comparison Table (Enclosure 68). 3.4-11 The plant does not have the RHR autoclosure portion of the RHR System interlock as the system was deleted from the plant design. However, the portion of the interlock which prevents the valves from opening when system pressure is in excess of the setpoint has been retained. As such, the note referring to the autoclosure interlock is deleted l from improved TS 3.4.14 Condition C and SR 3.4.14.2 and SR 3.4.14.2 is modified to be consistent with LCO 3.4.12. SR 3.4.14.3 is not used. 3.4 12 Not applicable to Callaway. See Conversion Comparison Table (Enclosure 68). 3.4 13 In conformance with the current TS. the RHR Isolation Valves which are RCS PIVs are excluded from being retested following flow through the valves. The remainder of the PIVs are check valves and therefore the requirement has been reworded to apply to flow through check valves only. 3.4 14 Revises grouping of required detectors to be consistent with the curre'nt TS and with the plants' licensing basis on Reg. Guide 1.45 Regulatory Position C.3. 3.4 15 LC0 3.0.4 is applicable to Condition [C] of ITS 3.4.15 (required containment atmosphere gaseous radioactivity monitor and containment cooler condensate monitoring system inoperable) because other mechanisms (i.e. , containment atmosphere samples, RCS inventory l balance, containment sump monitor, etc.) exist which are capable of l adequately detecting RCS leakage and because a 30 day A0T is usually I accompanied by an LCO 3.0.4 exception (e.g., PAM and Remote Shutdown Technical Specifications). Therefore, an exception to LC0 3.0.4 was added to Condition [C]. As there is already an LC0 3.0.4 exception to JUSTIFICATION FOR DIFFERENCES TS 2 5/15/97

4 CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431 Page 1 of 7 SECTION 3.4 DIFFERENCE FROM NUREG-1431 APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY 3.4-01 Clarifies intent of wording for the allowance to remove Yes Yes Yes Yes pimps from operation by changing "de energized" to

        " removed from operation." consistent with traveler TSTF-153.

3.4-02 This change revises Condition A of ITS 3.4.6 to cover any Yes Yes Yes Yes required loop's inoperability and adds REQUIPID ACTION A.2 indicating that cooldown to MODE 5 is only required if an RFR loop is OPERABLE. Condition B is deleted. 3.4-03 The current technical specifications allow I hour for the Yes Yes Yes Yes de-energization of all RFR punps. [] 3.4-04 The symbol ">" in DCPP LCO 3.4.8 Note 1.a is replaced by Yes No No No the words "at least." 3.4-05 This change is being made consistent with the current No. These Yes No - Similar No - Similar assupptions used in the analysis. The analysis credits operational analysis analysis three operational restrictions below 350'F. to ensure that restrictions are assumptions not assunptions not the reactor vessel is protected. not part of the contained in CTS. contained in CTS. current TS. 3.4-06 Plant specific safety analyses do not allow injection from Yes - DCPP analysis Yes - CPSES Yes - WCNOC Yes - Callaway safety injection punps but do allow CCP injection. assumes one CCP analysis assumes analysis assumes analysis assumes only. two CCPs. one CCP only. one CCP only. 3.4-07 The word "all" in the DCPP LCO 3.4.12 APPLICABILITY is Yes No No No replaced by the word "any." 3.4-08 The current licensing basis as contained in the Technical No - DCPP LTOP Yes Yes - See OL Yes - See OL Specifications requires performance of this surveillance design does not use Amendnent No. 49. Amendnent No. 42. on a frequency of 72 hours. PJR relief valves. 3.4-09 'h: pMt dx; =t h;..n.:. J ;'. ^JR oa.e. i=l:tte "; OCrr LTO -Ves--A//j) -Ves-A/[/)

                                                                                                                -Ves-A/[/)

9. i.. = ?^ E = = E 3 M.. M.. 9,,, _9 7_ E....-

                                                              . 9. M. . P.. . .. .=.

an a . , . u . > . A/,-/- uee/, d9 CONVERSION COMPARISON TABLE - NUREG-1431 5/15/97

l ADDITIONAL INFORMATION COVER SHEET l ADDITIONAL INFORMATION NO: O 3.4.12-4 APPLICABILITY: CA, CP, WC l l REQUEST: ITS Bases 3.4.12 Applicability (Comanche Peak, Wolf Creek, and Callaway) l l Comment: The intent of the addition to the end of the first paragraph of the l Applicability Bases is unclear. The LCO applies if the head is on. The added discussion essentially states LTOP (COMS) protection is not needed with the head on l and the bolts fully detensioned. If that is the argument then rather than adding it to the Bases discussion, the case should be made for modifying the LCO Applicability. FLOG RESPONSE: This comment is not applicable to Comanche Peak as this information was not in the CPSES ITS Bases. The Applicability for ITS LCO 3.4.12 includes MODE 6 when , the reactor vessel head is on. With no fuel in the reactor vessel the plant is not in MODE 6. ' The statement was placed in the iTS Bases to indicate that low temperature overpressure protection (LCO 3.4.12) is not required to be OPERABLE with no fuel in the reactor vessel. , There may be some plant conditions when the reactor is defueled that warrant placing the I l reactor vessel head on the vessel for radiological concerns. In these situations, the ! requirements of LCO 3.4.12 are not required to be met. The inserted Bases words are being deleted and plant procedures will provide the appropriate guidance for the plant conditions when no fuel is in the reactor vessel. i l l ATTACHED PAGES: Attachment 10, CTS 3/4.4 - ITS 3.4 Enclosure 58, page B 3.4-73 l l I l

L40" Sy:;tcm B 3.4.12 l _ BASES l LC0 g E- r..v.i s; x ;c H.. .o. . .

                                                                                                        . .o. < a .c      ..

1 (continued)

                                                                        ..o:
                                                                                 .w.:.).4            - .. ..        ;.,,7.:.......s.12.
                                                                                                                                   .               .:.'v.

f .t.c : . t . . (. . m.. , , .. . . .y . . , . . w .

                                                                                                                            ...     .g;, . c. , . ., .: .. , . i
                                                - ,; ;. g. . n . q.

g s- ,..:p. 4. e : ji n y, ..c.. . p 33 : :.; p aw.y a. i.7 m E k A depressurized RCS and an RCS vent. An RCS vent is OPERABLE when open with an area of x . E-07 square inches. Each of these methods of overpressure prevention is capable of mitigating the limiting ETOP % . transient. l redUmb APPLICABILITY This LCO is applicable in M00E14_ ~ any RCS cold leg temperature is s 0758F @ qn MODE 4) in MODE 5, and in HODE 6 when the reactor vessel head is on. The pressurizer h 7Ab/ safety valves provide overpressure protection that meets the Reference 1 P/T limits . : . -

                                                                                                                       ".        h a 275"T. When the reactor vessel head is off, overpressurization cannot occur.
                                                                                                                                                                        .7.4./ N a      _w__1_
                                                  .eo.,
                                                                                                         -..-n.

_ _ _ _ . w. - ._._. _ ___ _ _ _ _ _ _ . _ . . g - . . . . _ h -nstarrumimmwevrmaamammacs BWafBltBS

       ,   x..

(continued) MARK UP OF NUREG 1431 BASES B 3.4 73 5/15/97

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.4.12-5 APPLICABILITY: CA, WC ; I REQUEST: Differences 3.4-18 and 3.4-45 (Wolf Creek and Callaway) Comment: The justification for the 4-hour pump swap is inadequate. The STS allows 15 minutes. The CTS is used as justification however, finding a pump inoperable and then restoring it (which is the case covered by the CTS) is very different than simply switching from one operable pump to another. FLOG RESPONSE: Four hours is a reasonable time restriction for swapping centrifugal charging pumps (CCP) during the low temperature overpressure protection (LTOP)/ cold overpressure mitigation system (COMS) Applicability. Current Technical Specification (CTS) 3/4.4.9 Bases state " Operation below 350 F but greater than 325 F with all centrifugal charging and Safety injection pumps OPERABLE is allowed for up to 4 hours.. .Given the short time duration that this condition is allowed and the low probability of a single failure causing an overpressure event during this time, the single failure of a PORV is not assumed. Initiation of both trains of Safety injection during this 4-hour time ! frame due to operator error or a single failure occurring during testing of a redundant channel are not considered to be credible accidents." Additionally, CTS 3.5.4 requires all Safety injection pumps and one CCP to be inoperable. If this requirement is not met, then four hours is allowed to retum the pump (s) to an inoperable status. Performing CCP swap operations for maintenance activities requires both pumps to be capable of injecting for a limited period of time. During the time allowed for pump swap operation, the inoperable CCP must first be restored to OPERABLE status to meet ITS LCO 3.5.3 (MODE 4) and USAR/FSAR Section 16.1.2.3 (one OPERABLE CCP in boration flow path, MODES 4-6). Then the other CCP must be rendered incapable of injecting. In order to render the other CCP incapable of injecting into the RCS, the requirements of ITS SR 3.4.12.2 must be met. SR 3.4.12.2 Bases states that a pump is rendered incapable of injecting into the RCS through removing the power from the pumps by racking the breakers out under administrative controls. The Bases also state that an altemate method of cold overpressure protection control may be employed using at least two independent means to render a pump incapable of injecting. Each method includes local actions (e.g., breaker racked out and tagged, valve closed and tagged). These actions for restoring the one CCP and then rendering the other CCP incapable of injecting into the RCS cannot be performed from the control room. Swapping of CCP trains is a short duration evolution but must be performed in a controlled manner especially when coordinating activities outside the control room. The 4 hour time allowance provides for normal operation of the piant and allows piant manipulations / evolutions to be performed in a time frame in which they can be safely performed. Amendment No.103 (Callaway) and Amendment No. 89 (Wolf Creek) revised current TS 3.5.4 to provide a 4 hour AOT to restore one CCP to an inoperable status in MODES 5 and 6. This 4 hour AOT was specifically reviewed and approved by the NRC as noted in their safety evaluations for those license amendments. This portion of the COMS/LTOP Applicability is the most limiting, as it may involve water solid operation. Current TS 3.5.3 (SR 4.5.3.2) allows 4 hours to secure one CCP after entering MODE 4 from MODE 3. Current TS 3.5.2 requires both

I l i CCPs to be operable in MODE 3. Therefore, all of the ITS 3.4.12 Applicability is based on the '

         - current TS except for MODE 4 beyond 4 hours after entry from MODE 3. NSHC LS-24 justifies 4 hours for all of MODE 4.

ATTACHED PAGES: i None i l l l l l i I l i-l l-l l t e l-l l

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: Q 3.4.13-2 APPLICABILITY: CA, CP, DC, WC REQUEST: Change 6-26 LS 30 and Difference 3.4-36 (Diablo Canyon, Callaway and Wo'f Creek) Comment: TSTF-116 has not yet been approved by the NRC. FLOG RESPONSE: TSTF-116, Rev. 2 is currently under NRC review. This change provides assurance that the RCS water inventory balance will provide meaningful results. The proposed wording in TSTF-116, Rev. 2, was modified from TSTF-116, Rev.1, and these modifications have been incorporated into the ITS. The FLOG continues to pursue the changes proposed by this traveler. This Comment is also applicable to CPSES based on the applicability of JFD 3.4-36. ATTACHED PAGES: Attachment 10, CTS 3/4.4 - ITS 3.4 Enclosure 5A, Traveler Status Sheet ! Enclosure SB, page B 3.4-87 Enclosure 6A, page 6

l I l 1 l

IlOUSTRY TRAVELERS' APPLICABLE TO SECTION 3.4 ! 4 TRAVELER # STATUS DIFFERENCE # C0 K NTS 4 TSTF 26 Incorporated 3.4 32 - Approved by NRC. TSTF 27. Rev. [ Incorporated 3.4 33 A p / l A/4 y c. ((f jg I TSTF 28 Incorporated 3.4 22 Approved by NRC. TSTF 54. Rev.1 Incorporated M $$sY37 chid [$nly. W.s.4 49

                                                                                                                                              .t   1 TSTF 60                       Incorporated                                3.4 15       Approved by NRC.                             Q!

TSTF 61 Not Incorporated NA Minor change that is adequately addressed in the Bases. 1 TSTF 87, Rev. f Incorporated 3.4 31 Affrwe/ lyNA6 - 7At.3,4y ; TSTF 93, Rev.$ Incorporated 3.4 17 M$r"oIM/orEia'way in OL ,. L3 4 7,f,4-3 Amendment No.105. g-zf x3 i: TSTF 94;/ev. I Not Incorporated NA Retained current TS. yg-y.4( 4

          ;= 1R f=. !                  k -,,~,. %                                   3.'-36                   g 3 4 f /    -et y) -c; TSTF 108. Rev.1               Not Incorpotated                               NA        LCO 3.4.19 does not apply.                   E, TSTF 113 Rev. [Yncorporated                                                3.4 39   d lr,4,/4-/[f)24//-3 3 21- 3/j $

TSTF 114 Incorporated NA ApprovedbyNRC: Bases 3.4.7 g - m __ changes only. _- , v.. TSTF 116, Rev. d~3orporated 3.4 36 , 67 74/3-P_ ) H 7.f. ,e/ / f TSTF 136 I $ ratea NA B[MI43 #aM N .I changes 2 only. ,, Tg-y44 'g TSTF 137 Incorporated NA B[M5Nchfd[$nly. rg .r.4 W' TSTF 138 Not Incorporated NA Inconsistent with RCS loops requirements of ITS 3.4.5 and

3.4.6.

C TSTF 151gev. / Incorporated NA Bases 3.4.11 changes only.7k-S.+tpf tf TSTF 153 Incorporated 3.4 01 A //'**V M k A/AC-- TX-5,fAPT TSTF 162 y Incorporated NA EN.9/chInhI>'only. 7X-34-g Incorporated g 43 23, 3.4 45, See also hs3'.i-[a$f 33I 20 . # T.cc n~ 3 'T Incorporated

          - ..                                                                      3.4 35                 g y,4//-p       o    y.f   jj_    g
    %                         ~$    -

Incorporated 3.4 10 DCPP only./)pr,w/lyNgd gy, ( WOG 87, gev. 2. Incorporated 3.4 47 $ I*//-f 7,0 7.f.'e MARK UP OF WOG STS REV 1 (NUREG 1431) 5/15/97

RCS Operational LEAKAGE B 3.4.13 BASES SURVEILUWCE SR 3.4.13.1 (continued) (f[/0wer/ eve /;.cM/e afru es,AoCyeestayepy

*/

REQUIREMENTS The RCS water inventory balance myntuw-...nf..; pep M* rd (Na "' steady state operating conditions} C ad x;r egretir; pr;;;;re. .pj g , Therefore, m this SR is not required d to be performed in "0000 c...d 4 until 12 hours of g f)f.f/N ! i m steady state operatioq x.c egr;tirs pres;;r; Mn b ;; est;bli kd. , ,- + L Ged/rna) Q 2. + Gen -l

;                                                            Steady state operation is r; quired m to perform a proper inventory balanceg calculations during m                                                  & J.+.6en-/          ;

..;auncir; i erc ne; a;;ful end e . .;e requires tt.; Surveillexc to bc x; Ma study st;te i; ;;t;bli;Md. For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and j makeup tank levels, makeup and letdown, and RCP seal injection
and return flows.

An early warning of pressure boundary LEAKAGE or unidentified 3 LEAKAGE is provided by the automatic systems that monitor the 1 containment atmosphere radioactivity and the containment sump

level. It should be noted that' LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. These leakage detection systems are specified in LC0 3.4.15. "RCS Leakage Detection Instrumentation."
The 72 hour Frequency is a reasonable interval to trend LEAKAGE: and recognizes the importance of early leakage detection in the prevention of accidents. A Ltc .r.dcr tk r,;;.;xj ;;1x ,

stete; tMt tt.is = is required te b; grierad durirs steady st;t; egretien. , SR 3.4.13.2 ~ This SR provides the means necessary to determine SG OPERABILITY in an operational MODE. The requirement to demonstrate SG tube

integrity in accordance with the Steam Generator Tube Surve111anc'e Program emphasizes the importance of SG tube

;                                                            integrity, even though this Surveillance cannot be performed at i                                                                                                                                          (continued)

MARK UP 0F NUREG 1431 BASES B 3.4 87 5/15/97

CHANGE NUPEER JUSTIFICAT10N

  ! , .-                but would limit the exception to prior to entering MODE 2. This change is consistent with traveler OCO. - 777F-air,                            g f. f,// -g.

3.4 36 SR 3.4.13.1 and ACTIONS for LC0 3.4.15 are revised with the addition of a note per traveler TSTF 116,."r;.1. The note addresses the concern $ f.f,/1-l1 that an RCS water inventory balance cannot be meaningfully performed , unless the unit is operating at or near steady state conditions. The ' Note added to the surveillance provides an exception for operation at less than steady state conditions. The RCS water inventory balance will only be allowed to be deferred for 12 hours after re establishing steady state conditions. 3.4 37' The primary to secondary leakage limits are revised per Callaway OL Amendment No. 116 dated October 1. 1996. 3.4 38 tent with traveler TSTF 105. the details on ovG Dy WhichI the RCS is verified are moved f 3.4.1.4 to the Bases. Moving this informatio ows the use of precision heat balances, elbow taps. r acc thods in order to perform this verificat nd is consistent with the - ev. 1 philo moving clarifying information and descriptive s_out he is to the Base u 4 faa 'lQ3'.+./-/ Onvenin %furiren~y Tille , Encleu j;g(, }, C, n i8 ' 3.4 39 The shutdown requirements of ITS .4.11 would requi the plant to reduce Tm to < 500'F within 12 hours, rather than go-to MODE 4. to address the concern of entering (COMS] LCO 3.4.12 Applicability with l inoperable PORVs. For consistency. the shutdown requirements of ITS 3.4.16 are a o revised to allow 12 hours to red e T_ to < 500*F. This change is c sistent with traveler TSTF 113 .fA/fEgPgA-f] 8 84//~d

                                                                 ^
                                                                                          ^

wF= =M2

3.4 40 Consistent with traveler W96-99-W'the Note t'o SR 3.4.1.4 would be & S.f:/ .2 modified to provide additional time to perform an RCS precision flow rate measurement. The time allowed would be changed from 24 hours to 7 days. This change is acceptable because other indication of RCS flow is available (SR 3.4.1.3. RCS total flow meters) and additional time normally would be required to establish plant conditions suitable for the precision heat balance. Since this parameter does not normally change significantly and the flow meters can be used in the interim, there'is no need to perform this SR within the 24 hour period specified in NUREG 1431 Rev. 1. The 7 day period provides sufficient time to j establish steady state plant thermohydraulic conditions and obtain ! equilibrium xenon. In addition. the THERMAL POWER specified in the j Note would be changed from the generic value in brackets (90% RTP) to j 95% RTP. This change is acceptable because it specifies a power level l in better agreement with current operating procedures for performing a precision heat balance. Current TS do not specify a power level for this measurement. 3.4 41 Not applicable to Callaway. See Conversion Comparison Table (Enclosure 68). JUSTIFICATION FOR DIFFERENCES TS 6 5/15/97

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.13-3 APPLICABILITY: CA, CP, WC REQUEST: ITS 3.4.13 Bases LCO c. (Wolf Creek, Callaway, and Comanche Peak) Comment: How is the addition of what does not constitute identified leakage consistent with the definition in ITS Section 1.17 FLOG RESPONSE: The three categories in the definition of identified leakage do not include , leakage outside containment. The added text in the ITS 3.4.13 LCO Bases on what does not constitute identified leakage is unnecessary and will be removed. ATTACHED PAGES: Attachment 10, CTS 3/4.4 - ITS 3.4 Enclosure SB, page B 3.4-84

RCS Operational LEAKAGE B 3.4.13 BASES (continued) LCO RCS operational LEAKAGE shall be limited to:

a. Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material deterioration. LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB @

                                                                                                                                 & 3.+.Gm-I LEAKAGE past seals 3 end gaskets                        .     +2            -
                                                                                                                                   .~

is not pressure boundary LEAKAGE. . b. Unidentified LEAKAGE One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this LCO could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary.

c. Identified LEAKAGE L'p to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of gidentified LEAKAGE and is well within the capability of the RCS Makeup System.

Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE). __g $.Z4./3-3 gr-- -n.- ,--- . .

                                                                                                                . ' ' ~. ' R rd_----                                                                                       5
                                          -- _ --                                                         ,-                  -g TP~~"' Violation of this LCO could result in continued degradation of a component or system.
 ; ~ f.

(continued) MARK UP OF NUREG 1431 BASES B 3.4 84 5/15/97

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.13-6 APPLICABILITY: CA REQUEST: ITS 3.4.13 Bases LCO a. (Callaway) Comment: The intent of the addition tiut hakage past instrumentation lines not being pressure boundary leakage is unclear. Is that leakage upstream of isolation valves? If it is, is there a line size limit and is this consistent with the description of pressure boundary in the FSAR and the definition in ITS Section 1.17 FLOG RESPONSE: This Bases change refers to 3/8 inch tubing for instrument connections to ASME Class 1 fluid piping downstream of the root valves and to 1/8 inch core exit thermocouple sheaths. These instrument lines are not part of the reactor coolant pressure boundary (RCPB), as discussed in FSAR Table 3.2-1 Notes (9) and (10). As further stated in Sub-article NCA-1130(c), the scope of ASME Section ll1 does not apply to instrument tubing and that tubing is not designed or specified to be part of the RCPB or provide a pressure retaining barrier. As discussed in FSAR Sections 9.3.4.2.3.5 and 15.6.5.2, normal charging can accommodate a 3/8 inch break and maintain normal pressurizer level such that the ECCS is not actuated. This Bases change does not refer to leakage upstream of instrument root valves. There is no conflict with ITS Section 1.1. ATTACHED PAGES: None

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.14-1 APPLICABILITY: CA, CP, WC l REQUEST: Difference 3.4-13 (Callaway, Wolf Creek and Comanche Peak) Comment: What is the justification for restricting the testing to check valves with the addition of the term " check"in three places in SR 3.14-1 [ sic] and its Bases? All PlVs , at a plant may be check valves however, the addition is not consistent with the "or l isolation valve" part of the first sentence of the SR Bases or with the words of required l Action A of ITS 3.4.14. For Callaway and Wolf Creek simple deletion (sic) of " check" causes a problem with CTS 4.4.6.2.2.d and 4.4.5.2.2.d for Comanche Peak. FLOG RESPONSE: CTS 4.4.5.2.2 for Comanche Peak requires surveillances be performed on each RCS PlV listed in Table 3.4-1. The valves listed in this table are not all check valves. All the valves listed are subject to the testing frequency of items SR 4.4.5.2.2.a, b, c, and e. In addition, testing of the check valves within 24 hours of actuation was specifically addressed in item d. This CTS surveillance does not contain a 24 hour test requirement for non-check valve PlVs. The STS equivalent of 4.4.5.2.2 for Comanche Peak is SR 3.4.14.1. However the STS SR does not appear to limit the 24 hour test requirement to check valves only. Therefore that portion of the STS surveillance was modified to be consistent with the CTS. The Bases were similarly modified. l CTS 4.4.6.2.2.d for Callaway and Wolf Creek is similar to CTS 4.4.5.2.2.d for Comanche l Peak. However, CTS 4.4.6.2.2.d for Callaway and Wolf Creek does not specify check valves ! only (as does the Comanche Peak counterpart). Nevertheless, all the valves subject to CTS l 4.4.6.2.2.d in Table 3.4-1 for Callaway and Wolf Creek are check valves, given that the CTS SR wording excludes the RHR suction isolation valves. It was decided that the STS wording i should be revised consistent with the wording for Comanche Peak to reference only check valves rather than bring forward the CTS list of PlVs and the RHR suction isolation valve exclusion. ATTACHED PAGES: Attachment 10, CTS 3/4.4 - ITS 3.4 l Enclosure SB, pages B 3.4-91 and B 3.4-93 1 l l l

RCS PIV Leakage B 3.4.14 BASES LCO suggests that something is operationally wrong and corrective (continued) action must be taken. The LCO PIV leakage limit is 0.5 gpu per nominal inch of valve size with a maximum limit of 5 gpe. Oc prais; criteria ;f 1 ,,, fer eli v;lv; ;in; 1;p;;d a ur.ju-tifi;d Freity en th; i lagr v;1;;; witte.t gaidir; lnfe atia = pt,.atiel velv; d, err.d;tien .r.d r;;ulted in hi#;r gi;. oral redi;tien ;,,p;;r;;. A ;tudy cercluid ; inhg r.t; li;;it bs;d en velv; ;in an

q-:rier t; a sir.;is alicu:iic veba.

Reference permits leakage testing at a lower pressure WO-S,f-4/O differential than between the specified maximum RCS pressure and the normal pressure of the connected system during RCS operation (the maxima pressure differential) in those types of valves in which the higher service pressure will tend to diminish the overall leakage channel opening. In such cases, the observed rate may be adjusted to the maximum pressure differential by assuming leakage is directly proportional to the pressure differential to the one half power. APPLICABILITY In MODES 1, 2, 3, and 4, this LCO applies because the PIV leakage potential is greatest when the RCS is pressurized. In MODE 4 valves in the Rm flow path are not required to meet the requirements of this LCO when in, or during the transition to or from, the R R mode of operation. In H0 DES 5 and 6, leakage limits are not provided because the lower reactor coolant pressure results in a reduced potential for leakage and for a LOCA outside the containment. ACTIONS The Actions are modified by two Notes. Note 1 provides clarification that each flow path allows separate entry into a Condition. This is allowed based upon the functional independence of the flow path. Note 2 requires an evaluation of affected systems if a PIV iginoperable. The leakage may have 1Mt 1-dir- T -- affected system operabilityg 1 ^ #gr dt- i^- ^' :$1",M

                                                                                                  . .. . ,/ ' eWM U./6/

9Yri 4 _, "_2.y.~-- 2 ;Z t;-- .ct:_ _ f'. 2 . Ll 7.2 _ m, _,,,_ _ u..f _ ,.4, 7y;17- , (continued) t MARK UP OF NUREG 1431 BASES B 3.4 91 5/15/97

RCS PIV Leakage B 3.4.14

l 1

BASES

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

~ also reduces the potential for a LOCA outside the containment. The allowed Completion Times are reasonable based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. 1 I The inoperability of the RM eutxle;ur; oeo .- J

interlock render; the "J: ", ;uction i;;l;tien velv;; iraep;bi; ef j i;eletin; in r;;pe a . t; ; high pre;;ur; eenditien end prs;ating inadvertent cpening of the valves at RCS pressures in excess of the RR systemt design pressure. If the RlR interlock is inoperable, j ;ute le;;r; j operation may continue as'long as the affected RHR suction
penetration is closed by at least one eis;;d ;. nuel er

] deactivated -automatie valve within 4 hours. This Action M4 +-489-j accomplishes the pu se of the - - eut ale;;r; furetien. l\ kremde mane l SURVEILthlCE SR 3.4.14.1 REQUIREMENTS , Performance of leakage testing on each RCS PIV 3$.$1 % fSS.f,/p./) i used to satisfy Required Action A.1 s.r.d "equired L.;n A. is l l required to verify that leakage is below the specified . limit and j to identify each leaking valve. The leakage limit of 0.5 gpa per i inch of nominal valve diameter up to 5 gF maximus applies to each valve. Leakage testing requires a stable pressure i condition. For the two PIVs in series, the leakage requirement applies to

                           -  each valve individually and not to the combined leakage across both valves. If the PIVs are not individually leakage tested, one valve may have failed completely and not be detected if the other valve in series meets the leakage requirement. In this situation, the protection provided by redundant valves would be lost.

Testing is to be performed every 3 months, a typical refueling cycle, if the plant does not go into MODE 5 for at least 7 days. 1 (continued) HARK UP OF NUREG 1431 BASES B 3.4 93 5/15/97

l ADDITIONAL INFORMATION COVER SHEET

 ' ADDITIONAL INFORMATION NO: Q 3.4.14-3                      APPLICABILITY: CA, CP, DC, WC l

RECUEST: ITS 3.4.14 Actions Notes 1 and 2 Comment: The adoption of the STS notes (especially #1 which is a less restrictive change) is not discussed / justified. FLOG RESPONSE: A new DOC (609-L5-38) has been added to include ITS 3.4.14 Action Note 1 to the CTS markup. This note allows separate Condition entry for each pressure isolation valve (PlV) flow path made inoperable by an inoperable PlV. Also new DOC 6-30-A is added to include ITS 3.4.14 Action Note 2 which specifies entry into applicable Conditions and Required Actions for systems made inoperable by an inoperable PlV. ATTACHED PAGES: Attachment 10, CTS 3/4.4 - ITS 3.4 Enclosure 2, page 3/4 4-19 Enclosure 3A, page 13 Enclosure 3B, page 9 Enclosure 4, pages 2,62, and 63 i l I l l i

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE, 6 a

l

b. I gpm UNIDENTIFIED LEAKAGE, >

l l .

c. 600 gpd total reactor-to-secondary leakage through all steam

                                                                                 -"-'-"'-'"and                                     [-dS-A              ":"

generators - ' '- ' ' ' ' h any one steam generator, 150 gallons per day throug [

d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, L R:::ter C::1::t Sy;t;; g-27-LG.

e. * ;~ per :fDC ;_ r C0"ROLLE0 LE.*?^IE :t :

2:25 i 20 ;;i;, ::f pr :: r: .

f. The leakage .from

i'i d ir Tdl

each Eshall be Reactor Coolant limited to 0.5 gpm perSystem nominal inchPressure /-07-LG If of valve size up to a maximum of 5 gpm, at a Reactor Coolant System f- /p.-Lg. l pressure of 2235 i 20 psig.8ut g d f-LS APPLICABILITY: MODES 1,2,3,and4.f ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDB 6 hours and in COLD SHUTDOWN within the following 30 hours. ?

b. With any Reactor Coolant System leakage greater then any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage i

rate to within limits within 4 hours or be in at least HOT STAND within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. e CNf6M A

(-/I-LS ) With any Reactor Coolant System Pressure Isolation Valve leakage Ai

c. th: 1::k;;; r:.t; t: .;ith' . N9 greater than' h;tr:, the ;rabove be. in atlimit,Yr:d least HOT::li;it:

STANDBY .;ith!within , h HUTDOWN within the following%hotr , next

RCE 6 hours

                                                      ;r::: r:and   Of in  1 q:::th:2 500 p:ig.                           70       & 4-M CoLb                                         Q Qs.+.M % \   g ya pj (NEM) ENM47'                        D th:2   2225 p:i; but ;rt:ter th r 150 pei; cre 2110:: d.
                    *T::t pr::: r:: 1:::                                                       ::tt:1 t::t ;rt::Cr: 2; t:                       (-/g 4 6 Ob;;; .;d 1;;.k;.;; ;h;11                    b; ;dj.=t:d f:r th:b: dir::tly pr:pertiend t ? :::cre
                                            .......ir.; th; 1;;k;;; i:

p; diff;r;r.:is.1 t; ih: ::: S:1 # per? r $~Cl-LS f.%,

                                           .;1~.NMR7~ g J                                                                                                                                            f-/2-A1
                            .# w           rNsaw c                                  3/4 4-19        f                _

amandment No. 66,116 CALLAWAY - UNy_Cer) ikon enby i.c a /foweSbr- eack flrv a r.+.M e Se Enh, forde- e fr%, ,,) jt A ,- ry , n

                                   ~  = d e ! ~,y   pr=RelyninopermilahupdAc%

fra ,)Ja, co n rV, y

l CHANGE

    ;          NUMER             HSBC     DESCRIPTION

( CT .. licensee controlled doc ce Q3.5'.5b2 CN 2 15 LG of Enclosu conversion package) and by the thr e position lance in TTS ! SR ..

                                            .rNCERr SA-/3                                                         & 3% l+-3 7-01              -

Not applicable to Callaway. See Conversion Comparison Table (Enclosure 38).

               ' 01 LS 16    This change, in conformance with NUREG 1431 Rev. 1, revises the appilcability of the specification to MODES 1, 2, or 3 with (T,y) 2 500*F. The change deletes the requirement to perform an isotopic analysis for Iodine every 4 hours in Modes 4 and 5 and in Mode 3 below 500*F, whenever the reactor coolant exceeds its Dose Equivalent I-131 limit. In addition, this change deletes the requirement to perform the once per 4 hour surveillance for Dose Equivalent I-131 in the event the gross specific

activity limit is exceeded, in accordance with industry traveler TSTF 28. The latter is an unnecessary
requirement since the ACTION requires the plant to exit l the LCO's revised Applicability. This change is acceptable as offsite release of radioactivity in the event of an SGTR is unlikely for operation below 500*F, as the saturation pressure of the reactor coolant is below I the lift pressure settings of the main steam safety and I: [SG atmospheric steam dump] valves. I 8 02 LS 17 This change, in conformance with NUREG 1431 Rev. 1, adds 3 an exception to LC0 3.0.4 when operating in ACTION a, which is not in the CTS. This would allow H0DE changes under conditions that the plant is anticipating a return to acceptable activity levels within the 48 hour A0T. i j This exception is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient specific activity excursions while the plant l

remains at, or proceeds to, power operation.

8 03 LS 18 This change, in conformance with NUREG 1431 Rev. 1,

! revises the sample frequency from 72 hours to 7 days for l performance of a gasuna isotopic analysis. The 7 day frequency is acceptable based on the low probability of a gross fuel failure occurring which would significantly

        %                                  alter the analysis results.

l

. 8 04 H Consistent with NUREG 1431 Rev. 1, the CTS requirement to i

measure Iodine including I 131, I 133 and I 135 is l l DESCRIPTION OF CHANGES TO CURRENT TS 13 5/15/97

INSERT 3A-13 Q 3.4.14-3 i 6-29 LS-38 Consistent with NUREG-1431, separate Condition entry is allowed for l each flow path with excessive leakage from RCS PlVs. Although this specification provides a limit on allowable PlV leakage rate, its main purpose i i is to prevent overpressure failure of the low pressure portions of connecting systems. The leakage limit is an indication that the PlVs between the RCS and the connecting systems are degraded or degrading. Each flow path is allowed separate Condition entry based upon the functional independence of the flow paths. .That is, the required actions to isolate the high pressure portions of the affected flow paths, in order to protect the connecting low pressure systems, are not affected by leaking PlVs in other flow paths. 6-30 A ITS 3.4.14 Action Note 2 which specifies entry into applicable Conditions and Required Actions for systems made inoperable by an inoperable PlV is added to the CTS. This is an administrative change because it only makes explicit a general requirement that is already implicit in the CTS, I m 4

                                                                                                                                                          ~
   ,_,a ,     , ,. -    --m -w ..      s, - -

4-29 yggr gg_9 CONVERSION CONPARISON TABLE - CURRENT TS 3/4.4 Page 9 of 13 g.g Yer Ver Ve r Yer O .X+. f4-3 TE b EC M f 7# APPLICABIITV I" O #M-1 DIAELO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY NUMER DESCRIPTION 6 26 The CTS surveillance requirement for performing an RCS w3ter Yes No - Already part Yes Yes LS-30 inventory balance is modified. to allow deferral of the water of the CPSES inventory balance such that it would be performed within current TS. 12 hours after achieving steady state conditions. 6 27 RCS leakage detection system descriptions are revised for No - Current No - Current Yes Yes A consistency with current TS LC0 3.3.3.1 and FSAR Sections systems are systems are 5.2.5.2.2 and 11.5.2.3.2.2 applicable. applicable. 6 28 The current TS definition of CONTROLLED LEAKAGE is deleted. No - Retaining No - Retaining Yes - Moved to Yes - Moved to LG The RCP seal water return flow limit is moved to a licensee current TS. current TS. USAR. FSAR. L controlled document. This change relocates the reactor coolant system chemistry No - Amendnent Yes - To be No - Amenchent 89 No - Amendnent 103 7-01 specification from the Technical Specifications to a 98/97 relocated to relocated to TRM. relocated to USAR relocated to FSAR R licensee controlled document. ECGS. Chapter 16. Chapter 16. This change revises the applicability of the specification Yes Yes Yes Yes 8 01 LS-16 to HDDES 1, 2 or 3 with (T,) a 500*F. The change deletes the requirement to perform an isotopic analysis for Iodine every 4 hours in MODES 4 and 5 and in MODE 3 below 500*F. whenever the reactor coolant exceeds its Dose Equivalent lodine or at any time the reactor coolant exceeds Gross Specific Activity limits. This change adds an exception to LCO 3.0.4 when operating in Yes Yes Yes Yes 8 02 LS-17 ACTION a. This would allow MODE changes under conditions that the plant is anticipating a return to acceptable activity levels within the 48 hour A0T. This change revises the sanple frequency from 72 hours to Yes Yes Yes Yes 8-03 LS 18 7 days for performance of a ganuta isotopic analysis. This change revises the measurement of I-131.1-133 and Yes Yes Yes Yes 8-04 M I-135 to Dose Equivalent 1-131. This change revises the performance of the surveillance of Yes Yes Yes Yes 8-05 LS-19 the specific activity of the RCS following a 15% power change to H0DE 1 only. 5/15/97 CONVERSION COMPARISON TABLE - CURRENT TS

INSERT 38-9 Q 3.4.14-3 6-29 LS-38 Separate Condition entry is allowed for each flow path with excessive leakage from RCS PlVs. I 6-30 A~ ITS 3.4.14 Action Note 2 which specifies entry into applicable Conditions and .. Required Actions for systems made inoperable by an inoperable PlV is l= added to the CTS. l i. o l ![ f i I j l i l i: l t

l l l- NO SIGNIFICANT HAZARDS CONSIDERATIONS (NSHC) l - CONTENTS y (continued) LS 29........ ......................................... 51 LS 30.......... ....................................... 53 LS 31.................................................. 55 LS 32.......................................Not Applicable LS 33.......................................Not Applicable LS 34.................................................. 58 LS -35. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Not Appl i cabl e LS-36.................................................. 60 l LS 37. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Not appl i cabl e

                          / T-M                        -

t :1 & 2.4:/+-3 V. Recurring No Significant Hazards Considerations "TR" TR2...................................................e-4+ TR - 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 4 1 l l I I ? 2 5/15/97

_. . . -- - _ - _ _ . . . .=_ ._. __ _ l l IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATIONS l l l NSHC LS-38. $ f,f /f -? ! 10 CFR 50.92 EVALUATION l FOR i l TECHNICAL CHANGES THAT IMPOSE LESS RESTRICTIVE l REQUIREMENTS WITHIN THE TECHNICAL SPECIFICATIONS l 1 l l Consistent with NUREG-1431, separate Condition entry is allowed for each flow path with excessive leakage from RCS PlVs. Although this specification provides a limit on allowable PlV leakage rate, its main purpose is to prevent overpressure failure of the low pressure portions of connecting systems. The leakage limit is an indication that the PlVs between the RCS and the connecting systems are degraded or degrading. Each flow path is allowed separate Condition entry based upon the functionalindependence of the flow paths. That is, the required actions to isolate the high pressure portions of the affected flow paths, in order to protect the connecting low pressure systems, are not affected by leaking PlVs in other flow paths. This proposed TS change has been evaluated and it has been determined that it involves no l significant hazards consideration. This determination has been performed in accordance with , the criteria set forth in 10 CFR 50.92(c) as quoted below: i l "The Commission may make a final determination, pursuant to the procedures in 50.91, that a proposed amendment to an operating license for a facility licensed under 50.21 (b) or 50.22 or for a testing facility involves no significant hazards consideration, if operation of the facility in accordance with the proposed amendment would not: l

1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or

3. Involve a significant reduction in a margin of safety."

The following evaluation is provided for the three categories of the significant hazards consideration standards:

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

The proposed change adds a relaxation to the LCO by allowing separate Condition entry for each PlV flow path. Although this specification provides a limit on allowable l PlV leakage rate, its main purpose is to prevent overpressure failure of the low l pressure portions of connecting systems. The leakage limit is an indication that the l PlVs between the RCS and the connecting systems are degraded or degrading. Each flow path is allowed separate Condition entry based upon the functionalindependence l of the flow paths. Thus, the required actions to isolate the high pressure portions of the affected flow paths, in order to protect the connecting low pressure systems, are not affected by other leaking PlVs. Therefore, the proposed change does not involve a i l

IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATIONS i NSHC LS-38 d M MO (continued) significant increase in the probability or consequences of an accident previously evaluated.

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

The only accidents that are potentially associated with this proposed change are those related to the potential for an interfacing systems LOCA causing a failure of the low pressure portion of a system outside of containment with the resulting escape of radioactive material. The Required Actions for this LCO provide for the isolation of the flow path with valves that meet the same leakage requirements as the PlVs and which must be within the RCPB or, for DCPP and CPSES, within the high pressure portion of the system. The protection provided for the low pressure system continues to be maintained and is independent of the actions required to protect other flow paths that may also be affected. This change does not introduce any new overpressure accidents and the existing analyses remain valid. Thus, the proposed change does not create the possibility of a new or different kind of accident from those previously evaluated.

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

The proposed change does not affect the acceptance criteria for any analyzed event. Overpressure protection of each affected low pressure system continues to be provided by leak tested isolation valves which are independent of other flow paths. The margin of safety established by the LCO remains unchanged. Thus, there is no reduction in the margin of safety from that previously established. NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Based on the above evaluation, it is concluded that the activities associated with NSHC "LS-38" resulting from the conversion to the improved TS format satisfy the no significant hazards consideration standards of 10 CFR 50.92(c); and accordingly, a no significant hazards consideration finding is justified. l l l l l

i ADDITIONAL INFORMATION COVER SHEET ; ADDITIONAL INFORMATION NO: Q 3.4.14-4 APPLICABILITY: CA, WC REQUEST: Change 6-24 M (Callaway and Wolf Creek) Comment: Cold shutdown rather than hot shutdown is more restrictive however, the discussion does not address the extension nf the time from 12 to 30 hours. i FLOG RESPONSE: DOC 6-24-M is revised to add the following:

 " CTS LCO 3.0.3 specifies the standard shutdown track Completion Times when Required            l Actions and Completion Times aren't met as MODE 3 within G hours, MODE 4 within 12 hours, and MODE 5 within 36 hours. This DOC changes the termination point of the shutdown track in CTS 3.4.6.2 ACTION c from the non-standard MODE 4 with RCS pressure less than 600 psig in 18 hours to the standard MODE 5 within 36 hours. The cumulative effect is a more restrictive change."

ATTACHED PAGES: Attachment 10, CTS 3/4.4 - ITS 3.4 Enclosure 3A, page 12 : I 1 4 1 l 1 i i i

CHANGE

NUPEER HSlE DESCRIPTION l RCS hot leg suction isolation valves from inadvertently opening when RCS pressure exceeds the interlock setpoint.

l Upon failure of the interlock, the current TS permits , continued operation for 72 hours for restoration of the - affected subsystem. The improved TS requires action within 4 hours to isolate the affected RHR subsystem. . Thus the new ACTION decreases the probability of an l-intersystem LOCA upon failure of the interlock. This is a , l more restrictive change and the new ACTION is in & LC0 3.4.14 Condition C of the improved TS. 7 e? 6 23 LS 25 Specification 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 'g ACTION statements. This change is consistent with Mjy NUREG 1431 Rev.1 and traveler TSTF 60 and is acceptable d because of the diverse means available to detect RCS 7

1eakage. M 2A-MA gQ 9A/R 'i' 6 24 - M ACTION c of Specification 3.4.6.2 (Operational Leakage) is revised for consistency with NUREG 1431 Rev.1 to require f going to Cold Shutdown rather than going to Hot Shutdown I with an RCS pressure less than 600 psig. Ttis is e n. . l i;atei;;h. akth;, i;4ie:.:;r.t. 12V4ERT 34-/.38 03.f. l+- Q , PI 6 25 LS 26 Not applicable to Ca11away. See Conversion Comparison ; l Table (Enclosure 38). t 6 26 LS 30 The CTS surveillance requirement for performing an RCS water inventory balance is modified to allow deferral of the water inventory balance such that it would be [ performed within 12 hours after achieving steady state J conditions. The RCS water inventory balance must be 4 performed with the reactor at steady state conditions as b discussed in the ITS Bases. This change is in conformance , with traveler TSTF-116 Rev. 1. 6 27 A RCS leakage detection system descriptions are revised for consistency with current TS LC0 3.3.3.1 and FSAR Sections 1 5.2.5.2.2 and 11.5.2.3.2.2 { y 6-28 LG rrent TS definition of CONTROLLED L eted p

p' to be co nt with NUREG 143 . . The RCP seal j:p water rcturn fl moved to a licensee controlled documen S Jection - tions are established by tpa ow balance test procedur

                                       .rh/ reft ~ 38 ~l30                                       Q T.5~.5~~ 2 DESCRIPTION OF CilANGES TO CURRENT TS         12                                    5/15/97

1 INSERT 3A-128 Q 3.4.14-4 l CTS LCO 3.0.3 specifies the standard shutdown track Completion Times when Required 1 Actions and Completion Times aren't met as MODE 3 within 6 hours, MODE 4 within 12 hours, j and MODE 5 within 36 hours. This DOC changes the termination point of the shutdown track in 1 CTS 3.4.6.2 ACTION c from the non-standard MODE 4 with RCS pressure less than 600 psig in 18 hours to the standard MODE 5 within 36 hours. The cumulative effect is a more restrictive change. , l 1 l 1 f

l l 1 ADDITIONAL INFORMATION COVER SHEET i ADDITIONAL INFORMATION NO: Q 3.4.15-1 APPLICABILITY: CA, DC, WC , 1 REQUEST: ITS 3.4.15 and Bases ITS 3.4.15 Required Action E.1 (Callaway, Diablo Canyon and Wolf Creek) Comment: Callaway and Wolf Creek: As written ITS 3.4.15 does not implement CTS 3.4.6.1 as marked up (allowing up to two methods to be inoperable). Specifically, in the l ITS as written, with two monitoring methods inoperable TS 3.0.3 would have to be entered as there is no Condition for two methods inoperable. Diablo Canyon: ITS l 3.4.15 and Bases ITS 3.4.15 Required Action E.1. E.1 Bases state that "With two of I the three groups of leak detection monitoring not operable, the two groups will enter their respective ACTION and Completion statements." What in the construction of the ' ITS supports that statement and more importantly what is the justification for this as the CTS requires 2 of 3 groups of equipment to be operable? I FLOG RESPONSE: Given the independence of the three monitoring systems, the plant can simultaneously be in Conditions A and B, A and C, or 8 and C, but not in all three given Condition E invoking ITS LCO 3.0.3 if all three monitoring systems are inoperable. If ITS LOC l 3.0.3 were intended for the simultaneous inoperability of two systems, Condition E (Condition F in NUREG-1431) would be so worded. This is also supported by ITS page 1.3-1 (second paragraph of the Description section) and Example 1.3-3 which clearly state the plant may oe in multiple Conditions at the same time. DOC 6-23-LS-25 is revised to add:

  "In addition, given the leakage detection diversity, the ACTION for CTS 3.4.6.1 is revised to allow continued operation for up to 30 days under compensatory actions as long as one of the detection systems is OPERABLE. This condition is superior from a plant safety perspective l  than imposing a plant shutdown transient under LCO 3.0.3 which could give rise to an initiating l  event when the plant's leakage monitoring capability is degraded."

For Diablo Canyon, the Bases statement "With two of the three groups of leak detection monitoring not operable, the two groups will enter their respective ACTION and Completion statements" is deleted. ATTACHED PAGES: l l' Attachment 10, CTS 3/4.4 - ITS 3.4 l Enclosure 3A, page 12 1 Enclosure 3B, page 3 Enclosure 4, pages 49 and 50 1 i i i

l CHANGE EMER HSBC DESCRIPTION RCS hot leg suction isolation valves from inadvertently opening when RCS pressure exceeds the interlock setpoint. Upon failure of the interlock, the current TS permits %

continued operation for 72 hours for restoration of the affected subsystem. The improved TS requires action ~ within 4 hours to isolate the affected RHR subsystem. ( Thus the new ACTION decreases the probability of an l intersystem LOCA upon failure of the interlock. This is a e more restrictive change and the new ACTION is ir. Q LC0 3.4.14 Condition C of the improved TS. ~.' 3 6 23 LS '" Specification 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 4 ACTION statements. This change is consistent with % NUREG 1431 Rev. I and traveler TSTF 60 dnd is acceptable k because of the diverse means available to detect RCS T 1eakage. .[A/SF./1~ 2A-fGA & $A-lH %l 6 24 - M ACTION c of Specification 3.4.6.2 (Operational Leakage) is revised for consistency with NUREG-1431 Rev.1 to require [c going to Cold Shutdown rather than going to Hot Shutdown t with an RCS pressure less than 600 psig. Tl,i s i s e - -

                                        ;at.- 1 J a. E,M,,,r, e;+ieau..t. .ZN4 sty S&-/DB &3.f. l+- Q :

du 6 25 LS 26 Not applicable to Callaway. See Conversion Comparison [ Table (Enclosure 38). ; 6 26 LS 30 The CTS surveillance requirement for performing an RCS b water inventory balance is modified to allow deferral of the water inventory balance such that it would be performed within 12 hours after achieving steady state 1 conditions. The RCS water inventory balance must be i performed with the reactor at steady state conditions as R discussed in the ITS Bases. This change is in conformance %1 ' with traveler TSTF 116 Rev.1. 6 27 A RCS leakage detection system descriptions are revised for r j consistency with current TS LC0 3.3.3.1 and FSAR Sections - 5.2.5.2.2 and 11.5.2.3.2.2 m i w 6 28 LG rrent TS definition of CONTROLLED L eted y g to be co nt with NUREG 143 . . The RCP seal' W water return fl moved to a licensee controlled document. Jection tions are established by the ow balance test procedur ed from

                                        .rAlrEW SA -/2C                                              O .T. S'. C-:.2.        ,
      -DESCRIPTION _0F CHANGES TO CURRENT TS            12                                         5/15/97

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INSERT 3A-12A Q 3.4.15-1 In addition, given the leakage detection diversity, the ACTION for CTS 3.4.6.1 is revised to allow continued operation for up to 30 days under compensatory actions as long as one of the detection systems is OPERABLE. This condition is superior from a plant safety perspective than imposing a plant shutdown transient under LCO 3.0.3 which could give rise to an initiating , event when the plant's leakage monitoring capability is degraded. l l

e, CONVERSION COMPARISON TABLE - CURRENT TS 3/4.4 Pa 8 of 13 TECH SPEC CHANGE , APPLICABILITY NUEER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY 6-18 This change relaxes the requirement for PIV testing No - Not part of Yes Yes No - Already in LS-15 following operation in MODE 5. The previous requirement was current DCPP TS. current TS per testing following 72 hours in MODE 5 which is revised to Amendment 105. 7 days in MODE 5. This change removes the specific requirement for performing Yes Yes Yes Yes 6-19 TR-3 the PIV surveillance prior to returning a valve to service following maintenance, repair or replacement. 6-20 IST requirements are moved to Section 5 of the improved TS. Yes Yes No - WCGS does not No - Callaway does A have this not have this requirement. requirement. This change increases the RCP seal injection flow Coupletion Yes Yes No - See CN No - See CN 6-21 Time from 4 to 72 hours, with a new added verification that 6-28-LG. 6-28-LG. LS-35 at least 100% of the assumed charging flow remains available. 6-22 This change adds a new ACTION to isolate the affected FdR No - Not part of Yes Yes Yes M penetration within 4 hours if the RIR suction isolation current DCPP TS. valve interlock function is inoperable. The leakage detection system specification is revised such Yes No - The non- Yes Yes 6-23 LS-25 that the provisions of 3.0.4 are not applicabl ad/w, applicability of 41' ggg Y I YyCfN

                                                   *'N" C4h $* f y n /gwo'& &                                    7         g
                               'hvek'Iy / 003.0.3-                                                         TS.

6-24 Revises ACTION to require going to COLD SHUTDOWN rather than No - Not part of No - The 600 psig Yes Yes M HOT SHUTDOWN with an RCS pressure less than 600 psig. current DCPP TS. action is not part of the current TS. The Operation Leakage LCO has been modified to change Yes No - Leakage limit Yes No - Already part 6-25 allowed limit for RCS pressure isolation valves. of s 0.5 gpm is of current TS per LS-26 already part of Amendment 66. current TS. CONVERSION COMPARISON TABLE - CURRENT TS 5/15/97

IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATIONS 1 NSHC LS 25 10 CFR 50.92 EVALUATION FOR TECHNICAL CHANGES THAT IMPOSE LESS RESTRICTIVE REQUIREMENTS WITHIN THE TECHNICAL SPECIFICATIONS j i Specification 3.4.6.1 is revised such that the provisions of Specification 3.0.4 are I 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 Rev.1 and traveler TSTF 60 and is acceptable because of the diverse means available to detect RCS leakageg gJ.f/r_/ ZNMAYLC-2C This proposed TS change has been evaluated and it has been determined that it involves no significant hazards consideration. This determination has been

          . performed in accordance with the criteria set forth in 10 CFR 50.92(c) as quoted below:

1 "The Commission may make a final determination, pursuant to the procedures in 50.91, that a proposed amendment to an operating license for a facility licensed under 50.21(b) or 50.22 or for a testing facility involves no significant hazards consideration, if operation of the facility in accordance

                     .with the proposed amendment would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or

                   , 3.       Involve a significant reduction in a margin of safety."

The following evaluation is provided for the three categories of the significant hazards consideration standards:

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

Overall protection system performance will remain within the bounds of the previously performed accident analyses since no hardware changes are proposed. The primary function of the Leakage Detection Systems is to detect significant reactor ccolant pressure boundary (RCPB) degradation as soon as practical to minimize the potential for propagation to a gross failure. The TS requires multiple diverse systems to ensure that leakage from a variety of locations and leakage rates can be detected in sufficient time to take measures to place the plant in a safe condition. These systems are passive and can not initiate or increase the consequences of an accident, No credit is explicitly taken for these systems in the accident analyses. Entry into the applicable MOD or zwhile subject to the compensatory measures called for in the ACTION statemen sg will not have any effect on the status of the RCPB. The proposed change will f Y " O M kn Y N h onih Aq ryr/s m e fu ,rer., le ~$ve yy -fa 20 NO SIUIFICANT HAZARDS CONSIDERNLUM 4v s ___ 5/15/97

L INSERT LS-25 Q 3.4.15-1

                                                              .                                 l In addition, given the leakage detection diversity, the ACTION for CTS 3.4.6.1 is revised to allow continued operation for up to 30 days under compensatory actions as long as one of the l detection systems is OPERABLE. Allowing continued plant operation for 30 days, with           ,

l attemate methods invoked to further monitor RCPB integrity and at least one monitoring l system available to detect abnormal leakage, is superior from a plant safety perspective than imposing a plant shutdown transient under LCO 3.0.3 which could give rise to an initiating event when the plant's leakage monitoring capability is degraded. l l l 3 l

l IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATIONS i NSHC LS 25 (continued) , not affect the probability of any event initiators nor will the proposed change affect .the ability of any safety related equipment to perform its intended function. There will be no degradation in the performance of nor an increase in the number of challenges imposed on safety related equipment assumed _to function during an accident situation. Therefore, the proposed l change does not involve a significant increase in the probability or consequences of an accident previously evaluated. I 2. Does the change create the possibility of a new or different kind of

accident from any ccident previously evaluated?

rtVN There are no frdware changes nor are there any changes in the method by whichl any safety related plant system performs its safety function. The method of plant operatio i is unaffected si e the change is based on the acceptability , of the r .;.c.? ACTION statement n providing compensatory RCPB leakage d I fd G / l detection capability. No new accident scenarios, transient precursors, I failure mechanisms, or limiting single failures are introduced as a result of ! this change. Therefore, the proposed change does not create the possibility

                        .of a new or different kind of accident from any previously evaluated.

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

The proposed change does not affect the acceptance criteria for any analyzed ) event. There will be no effect on the manner in which safety limits or limiting safety system settings are determined nor will there be any effect on l those plant systems necessary to assure the accomolishment of protection

                      , functions. Thee will be no impact on any margin of safety.

NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Based on the above evaluation, it is concluded that the activities associated with NSHC "LS 25" resulting from the conversion to the improved TS format satisfy the no significant hazards consideration standards of 10 CFR 50.92(c): and accordingly, a no significant hazards consideration finding is justified. N0 SIGNIFICANT HAZARDS CONSIDERATION 50 5/15/97

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.15-2 APPLICABILITY: CA, WC REQUEST: CTS 3.4.6.1 b&c and CTS 4.4.6.1 b&c markups (Callaway and Wolf Creek) Comment: Have the systems been renamed, were the names in the CTS incorrect, or are different systems being relied on in the ITS7 FLOG RESPONSE: The same systems are being used. The system names in the CTS have been revised to be consistent with the names used in FSAR Section 5.2.5.2.2. DOC 6-27-A already provides this explanation. ATTACHED PAGES: None j l l I

l l l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 3.4.16-1 APPLICABILITY: CA, CP, DC, WC REQUEST: Difference 3.4-39 Comment: TSTF-113 has not yet been approved by the NRC staff. FLOG RESPONSE: See the response to Comment Number 3.4.11-3. l l ATTACHED PAGES: ( None 1 l i l l

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: CA 3.4-002 APPLICABILITY: CA REQUEST: Typographical errors are corrected. ATTACHED PAGES: Attachment 10, CTS 3/4.4 - ITS 3.4 Enclosure SA, page 3.4-40 Enclosure 58, page B 3.4-71 Attachment 19, Smooth ITS page 3.4-39 Attachment 20, Smooth ITS Bases pages B 3.4-66 and B 3.4-85 4 i 't V

RCS Specific Activity . 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.16 The specific activity of the reactor coolant shall be within limits. APPLICABILITY: MODES 1 and 2 MODE 3 with RCS average temperature (T,) :e 500*F. ACTIONS l l CONDITION REQUIRED ACTION COMPLETION TIME A. DOSE EQUIVALENT I 131 - NOTE - - -

                     > 1.0 C1/gs.                               LCO 3.0.4 is not applicable.

A.1 Verify DOSE EQUIVALENT I 131 Once per 4 hours within the acceptable region of Figure 3.4.161. 8 A.2 Restore DOSE EQUIVALENT 48 hours

                   >                                            I 131 to within limit.

l l B. Gross specific activity 0.1 N. 'ere Z 3.4.10.2. 4 hears v3'4 22e of the reactor coolant 7.et withir, li;;;it. E l 6 ::3.4-25 t ! B.Bg Be in MODE 3. w44-T,,, 6 hours j

500'F. --3.4 39; )

E g monumme h m a3 4-39: CA-1.4-po1 (continued) MARK UP OF WOG STS REV 1 (NUREG 1431) 3.4 40 5/15/97

e.

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B 3.4.12 BASES APPLICABLE RCS Vent Performance SAFETY ANALYSES (continued) With the RCS depressurized, analyses show a vent size of M E47 square inches is capable of mitigating the m wi0 ell;;;d LT

                                                     ~ ' ;;;rprc;;urc transient. The capacity of a vent this size is greater than the flow of the limiting transient for the HOP M configuration, se llPI pu;;;p erd ? - F ,- ",                                                                          -w ti. - t i; OPERABLE, maintaining RCS pressure less than the maximum pressure on the P/T limit curve.

The RCS vent size will be re evaluated for compliance each time the P/T limit curves are revised based on the results of the vessel material surveillance. I The RCS vent is passive and is not subject to active failure. l m .. ,-. '; The L T Sy;ts satisfies Criterion 2 of tra TC P;1 icy I LCO This LCO requires that the LT Sy;te;;; ' is OPERABLE. The LT Sy;ts 4 ' is OPERABLE when the n+nhunt m . coolant 4 input and M pressure relief capabilities are OPERABLE. l Violation of this LC0 could lead to the loss of low temperature ! overpressure mitigation and violation of the Reference 1 limits i as a result of an operational transient. To limit the coolant input capability, the LCO requires ~ "

                                                        . , . .         ..    .z..        . . o .<         ,.     .....o     o.. g; ;;r: p;;;;p .. . . . , .
                                                      .......: .           .m   .. c ,   .: :, . c .;, .         .1  capable of injecting into the RCS and all accumulator ischarge isolation valves 3 closed and                                                                                        yjg*
                               "            N bilizedl                                           M accumulator pressure is greater than equal to the maxi                              RCS pressure for the existina RCS cold leg f Q temperature allowed in the P                                          ,
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y i;jy '! $ ' e so ps.s l . $ f u ! d'.l .?Mi 4 (continued) MARK UP OF NUREG 1431 BASES B 3.4 71 5/15/97

RCS Specific Activity 3.4.16 n 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.16 The specific activity of the reactor coolant shall be within limits. APPLICABILITY: H0 DES 1 and 2, MODE 3 with RCS average temperature (T ) a 500*F. ACTIONS CONDITION REWIRED ACTION COMPLETION TIME A. DOSE EWIVALENT I 131 - - - - - - NOTE-- -- -- -

                > 1.0 C1/gm.                   LCO 3.0.4 is not applicable.

A.1 Verify DOSE EWIVALENT I-131 Once per 4 hours within the acceptable region of Figure 3.4.16 1. M A.2 Restore DOSE EWIVALENT 48 hours

              ,                                       I-131 to within limit.

B. Gross specific activity B.1 Be in H0DE 3. 6 hours of the reactor coolant

                > 100/E C1/gm.                 E B.2           Reduce T,y to < 500*F.                 12 hours           CA-F.+-don.

(continued) CALLAWAY PLANT ITS 3.4-39 5/15/97

l CONS B 3.4.12 l BASES APPLICABLE RCS Vent Performance SAFETY ANALYSES (continued) With the RCS depressurized, analyses show a vent size of 2.0

square inches is capable of mitigating the limiting COMS transient. The capacity of a vent this size is greater than the flow of the limiting transient for the COMS configuration, one centrifugal charging pump OPERABLE, maintaining RCS pressure less

! than the maximum pressure on the P/T limit curve. The RCS vent size will be re evaluated for compliance each time the P/T limit curves are revised based on the results of the vessel material surveillance. The RCS vent is passive and is not subject to active failure. The COMS satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). l LCO This LC0 requires that the COMS is OPERABLE. The COMS is l OPERABLE when the maximum coolant input and minimum pressure relief capabilities are OPERABLE. Violation of this LCO could lead to the loss of low temperature overpressure mitigation and violation of the Reference 1 limits as a result of an operational transient. To limit the coolant input capability, the LC0 requires that a maxims of zero safety injection pumps and one centrifugal charging pump be capable of injccting into the RCS and all accumulator discharge isolation valves be closed and immobilized when accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed in the PTLR. bur-The LC0 is modified by-threeV Notes. Note 1 allows two CA-3'.4-00 2. centrifugal charging pumps to be capable of injecting into the RCS for s 4 hours for pmp swap operations. This provides the necessary allowance to perform the pump swap activity in a controlled manner and provides sufficient time to complete the

activities necessary to restore a maximum of one centrifugal l charging pmp to a status capable of injecting into the RCS.

l l This is accomplished by racking out the breaker for one pump or i employing two independent means to prevent a pump start in accordance with SR 3.4.12.2. (continued) I CALLAWAY PLANT ITS BASES B 3.4 66 5/15/97 t

RCS PIV Leakage l B 3.4.14 BASES g!{ILlM A.1 and A.2 (continued) g_g m.wu mentm a-A CTT0 Alf Required Action A.1 requires that the isolation with one valve must be perfonned within 4 hours. Four hours provides time to , reduce leakage in excess of the allowable limit and to isolate i the affected system if leakage cannot be reduced. The 4 hour Completion Time allows the actions and restricts the operation with leaking isolation valves. l Required Action A.2 specifies that the double isolation barrier of two valves be restored by restoring the RCS PIV to within , limits. The 72 hour Completion Time after exceeding the limit I allows for the restoration of the leaking PIV to OPERABLE status. l This time frame considers the time required to complete the Action and the low probability of a second valve failing during this time period, B.1 and B.2 If leakage cannot be reduced, the system isolated, or the other Required Actions accomplished, the plant must be brought to a l MODE in which the requirement does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours and H00E 5 within 36 hours. This Action may reduce the leakage and also reduces the potential for a LOCA outside the containment. The allowed Completion Times are reasonable based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. 1 O The inoperability of the RHR suction isolation valve interlock ] could allow inadvertent opening of the valves at RCS pressures in i excess of the RHR system design pressure. If the RHR suction I isolation valve interlock is inoperable, operation may continue as long as the affected RHR suction penetration is closed by at least one deactivated automatic valve within 4 hours. This Action accomplishes the purpose of the interlock. l l (continued) CALLAWAY PLANT ITS BASES B 3.4 85 5/15/97 1 l l

1 ADDITIONAL INFGRMATION COVER SHEET ADDITIONAL INFORMATION NO: CA 3.4-003 APPLICABILITY: CA REQUEST: Incorporate changes related to OL Amendment No.124 dated April 2,1998. ATTACHED PAGES: Attachment 10, CTS 3/4.4 - ITS 3.4 Enclosure 2, pages 3/4 4-30, 3/4 4-31, and 3/4 4-36 Enclosure 58, pages B 3.4-9, B 3.4-16, B 3.4-63, B 3.4-64, B 3.4-67, B 3.4-66, B 3.4-69, B 3.4-77, and B 3.4-81 l l 1 i I 1 I

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                                       . . i .'                       OPERATION                      T
                                                                                                                                                           . ' - i!                            !                                  . *                  +                                                 6                .
                                                                                                                                                                   -      f.                   .                                       .               i              ,

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      = 1500
                                                                ',         .'                       ,' ,' ' s'                  .'

p3 _. . . . i . ! 6 , , 4 r > ,

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v / i r , i U3 .... i . . , . i 4 , i , , , , , , t i . f i i i , 4 1250-o .,, 1 '

                                     .'.'                                                                                                                                                                            ACCEPTA31.E
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6 .. . . ' " + -' ' . . .

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                                                                                                                                                                                           .         .' a   i OPERATION                                                        .! '

i,... , ie . 4 . fi 6 6 i 4 . I ! .6

                                     .           . .                                                                                                                                                                           . . 6 l                                    ie              i , , .                      6                      .
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r' 4 e j.

   -           750 -- e.n...                                                                                                              ::                                                                                               '

85 urns ,,, , , ii F/It. r 4~ . . .. ..s. , , . . , , . o i , i ii, ,.,.

                                                               '~ " " ' .

i! 6 6 . i , 1

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     =                             .             .0             f f f. ,j. .,

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    =                          _             , , ,

250 - . ' ' ' ' .' '. ,,' J '

   =                               .
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                                                                                                                                                                                        . . ' . ' ' ' .' ' . . .' ' .' .'i i .' .' ' '
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                                   ....                              ,        TSES. - r ii                                 +             , .           , i i ,                          e i !                                            . .
                                                                                                                                                                                                                                                           ..i i                        . , , !                     .i,.
                                                            ,                                         ..                   .             . .                          >>                i . . ,

0 s, i.. . . . . . ,. . 60 100 150 200 '250 300 3$0 460 450 500 In/ dicated Temperatu e (Deg.F) FIGURE 3.4-3

                             'allaway Unit 1 Reactor Coolant System Cooldown Limitatiorg (Cooldown Rates of 0. 20. 40. 60 and 100*F/hr) Applicable for the Firbt 20 EFPY (With Margins for Instrumentation Errors) Incluces Vessel flange requirements of 170*F and 561 psig per 10 CFR 50. Appendix G CA LAWAY - UNIT 1                                                                                                                      3/4 4-31                                                                                                Amendment No. %-ts,124
              /

n b

               ?   N

h h \ A y

y Callaway COMS Maximum Allowable Selpoints c- 2m

                                                                    '                                                                                                                                                                 5 N

y . N

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2

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                                                                                                                                                                                                                                    =

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w x a

m teco q A _ x m 5=

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                &                       a
                                                                                                                                  =                                                            ==

8 . = = s? N

                                                                                                                                                                                                                           \
               &                                                         + High PORV Setpoint + Low PORV Selpoint                                                                                                            x,          !

I O RCPs fee TRC5 < $00F - \ b at 4 RCFs for TRCS >= leoF

                                                                                                                                                                                                                      's s&
                                                                                                                                                                                                                                   '\ ':

C! 5 D bi

i

RCS P/T Ltits B 3.4.3 8 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4.3 RCS Pressure and Temperature (P/T) Limits BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes, i These loads are introduced by startup (heatup) and shutdown l (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup , and cooldown, within the design assumptions and the stress limits l for cyclic operation.

The PTLR contains P/T limit curves for heatup, cooldown,

inservice leak and hydrostatic (ISUI) testing, and data for the l maximum rate of change of reactor coolant temperature (Ref.1).

Each P/T limit curve defines an acceptable region for normal ! operation. The usual use of the curves is operational guidance j during heatup or cooldown maneuvering when pressure and treerature indications are monitored and compared to the applicable curve to determine that operation is within the allowabb region. l The LCO establishes operating limits that provide a margin to brittle failure.of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). The vessel is the component most subject to brittle failure, and the LCO limits apply mainly to the vessel. The limits do not apply to the pressurizer, which has different design characteristics and operating functions. ! 10 CFR 50, Appendix G (Ref. 2), requires the establishment of P/T limits for specific material fracture toughness requirements of the RCPB materials. Reference 2 requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use g g Y

3 of the American Society of Mechanical Engineers (ASME) yode, Section III, Appendix G (Ref. 3). T/>ere m*I4/e om d/<ewel indeklin WCAA/4f']4, affavadyJ N/C S Callway(deer,

              .                       The neutron embrittlement effect on the mateeihl toughness is fod4, reflected by increasing the nil ductility e +va reference temperature (RTer) as exposure to neutron fluence increases.

J MARK UP OF NUREG 1431 BASES B 3.4 9 5/15/97

L RCS P/T t.icits B 3.4.3 s BASES REFERENCES 6. Regulatory Guide 1.99, Revision 2. Hay 1988. (continued)

7. ASE, Boiler and Pressure Vessel Code Section XI.

Appendix E. O

                                    !                                                                                           CA-2.k6g P.        h) cat t+r94, " Ca lie a L;; j y,              j 0"Id~n Lra:+ C -va, E.- Ns ,,,, / spe-,4;,                                             ,
                                                 % f / 41~1.

kine In ed A/o, /.2 4- h $cr/if yd en3h l-iceae N)#f--fs b+ed A,0 cit s 1997 i U 2 v MARX UP OF NUREG 1431 BASES B 3.4 16 5/15/97

        ,..+                                                                                                                                          LO Sy;ta     ~

_ , , , B 3.4.12 1 B 3.4 REACTOR C00LM SYSTEM (RCS) B 3.4.12 L;w T( ,-;retur; r C;;r r;;;;r; Pre;;;ticr. (LT) m System i& BASES BACKGROUW The L O Cy;t s - controls RCS pressure at low temperatures so the integrity of the reactor coolant pressure boundary (RCPB) is not compromised by violating the pressure and temperature (P/T) limits of 10 CFR 50, Appendix G (Ref. 1). The reactor vessel is the limiting RCPB component for demonstrating such protection. The PTLR provides the maximum allowable actuation logic setpoints for the power operated relief valves (PORVs) and the maximum RCS pressure for the existing RCS cold leg temperature during cooldown, shutdown, and heatup to meet the Reference 1 requirements during the MGP MODES; 4.r ajyru ,/ /7 Ntc i

                                                                     & C*//my4, des./2.                                                                        CA-2.+-es2 The reactor vessel material is less tough at low temperatures than at nomal operating temperature. As the vessel neutron exposure accumulates, the material toughness decreases and becomes less resistant to pressure stress at' low temperatures (Ref. 2). RCS pressure, therefore, is maintained low at low temperatures and is increased only as temperature is increased.

The potential for vessel overpressurization is most acute when the RCS is water solid, occurring only while shutdown: a pressure fluctuation can occur more quickly than an operator can react to relieve the condition. Exceeding the RCS P/T limits by a significant amount could cause brittle cracking of the reactor vessel. LC0 3.4.3, "RCS Pressure and Temperature (P/T) Limits," requires administrative control of RCS pressure and temperature during heatup and cooldown to prevent exceedJng the PTLR limits. This LCO provides RCS overpres(sure protection by havin coolant input capability and having adequate pressure relief capacity. Limiting coolant 1@ut capability requires ;11 but er;

                                                                     ,,,,,,y,..     ...    ... u .,
                                                                          .,    m-...-     -e.,.     ..
                                                                                                           . . - incapable of in.jection                              N d3 into the RCS and isolating the accumulators. The pressure relief capacity requires either two redundant RCS lief valves or a depressurized RCS and an RCS vent of suffi ent size. One RCS
                                                                                                                             .1A/IEM*Aom                         CA-M-003 pe 83.Hf (continued)

MARK UP OF NUREG 1431 BASES B 3.4 63 5/15/97

l L T Sy;t b l B 3.4.12 i 1 . l BASES BACKGROUND relief valve or the open RCS vent is the overpressure protection (continued) device that acts to terminate an increasing pressure event.

With minimum coolant input capability, the ability to provide core coolant addition is restricted. The LCO does not require the makeup control system deactivated or the cafety injection

(SI) actuation circuits blocked. Due to the lower pressures in [1 the L40P H0 DES and the expected core decay heat levels, the makeup system can provide adequate flow via the makeup control valve. If conditions. require the use of more than one HPf-ee

                                                       .            > charging pump for makeup in the event of loss of inventory, A%(pumps can be made available through manual CA-7 NdS e r#en Ne NCl #or- o 4Le, Ecos                                                     _

The L T Sy;t s - - for pressure relief consists of two PORVs i with reduced lift settings, or two residual heat removal (R}R) f. suction relief valves, or one PORV and one RHR suction relief [ valve, or a depressurized RCS and an RCS vent of sufficient size. J Two RCS relief valves are required for redundancy. One RCS ? i relief valve has adequate relieving capability to kg h i overpressurization for the required coolant input capability.

                                                                        ~                                                         -
                                                                                                                                           ' CA-3.+-ssg    y ie
                     +

C' 1 ee+,ng a rt AarJean an sa] ar ca his of an overfretrure k,)),a,./. u) .pa ,ej, j ,,., V { fgyeh J [og e tuner he fow from one cenh;fuy/ ej, PORV Requirements NC/. ~7'/, Ar,.m "e,y o, "CCf" re-Lr is +le b-Ps ly-,-e la}ad skca ,,,,b,,Y/f

                                                                                                                                    ]      n, ,,g .pjn ,
                                                                                                                                                    ,       ll 3
                                                                                                                                                           'i    l g 1.f 43                               As designed for N= sy=, each nmv 1s' signaled to '                                                          l open if the RCS pressure approaches a limit determined by the                                        f E40P W actuation logic. The L40P % actuation logic                                                   '

monitors both RCS temperature and RCS pressure and determines t when a condition not acceptable +n .

                                                                                                                       -- <          the PTLR limits is approached. The wide range RCS temperature indications are auctioneered to select the lowest temperature signal.

l The lowest temperature signal is processed through a function !

generator that calculates a pressure limit for that temperature. (

(continued) MARK UP OF NUREG 1431 BASES B 3.4 64 5/15/97 m.-

M . B 3.4.12 i I l BASES i l APPLICABLE the Reference 3 4 analyses to determine the impact of the change SAFETY ANALYSES on the E40P 2 : acceptance limits. , (continued) l Transients that are capable of overpressurizing the RCS are categorized as either mass or heat input transients, examples of ] which follow: j l Hass Inout Tvoe Transients 1

a. Inadvertent safety injection: or i

b. Charging / letdown flow mismatch Heat Inout Tvoe Transients

a. Inadvertent actuation of pressurizer heaters;

b. Loss of RHR cooling; or

c. Reactor coolant pump (RCP) startup with temperature asyimaetry within the RCS or between the RCS and steam generators.

The following are required - - - - - during the L40P M H0 DES to ensure that mass and heat input transients do not occur, which either of the L46P ' - overpressure protection means cannot handle: j,,,,,_,,,,, g y 4.,

a. Rendering ;11 M er.; ll"I Iand incapable of CA-2. M injectiori:

T .- (He NCfis a }so a varf.Lle clu * -Ns CD/h5/\fs

b. Deactivating the accumulator discharge isolation alves in their closed positions; and

c. Dinile.;ir.;; -

o+- start of an RCP if secondary temperature is more than E*F above primary temperature in any one loop. ' . . - W 1. - :au n LCO 3.4.6, "RCS Loops H00E 4," and LCO 3.4.7, "RCS Loops H00E 5. Loops Filled," provide this protection. (continued) MARK UP OF NUREG 1431 BASES B 3.4 67 5/15/97

k i i i M

   .                                                                                                                                                                                                                                       m . v,      ,.7             -

j B 3.4.12 d

}                             BASES 4

i 1 APPLICABLE Heat Inout Tvoe Transients (continued) SAFETY ANALYSES

                                                    . . , . . . . . . . . .                .....               1 : . . . , .

l . . . . . , . . . . , . . . , .

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                                                                                                                                                ..            .                                                          .o.                ..

l

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                                                        . . , . .a              . . . . . .. . . , , .                          .             .                 ,,        .. . . . .                        u < .. ,..                                        ,

i [. . L. ' # .. .- ,,

                                                                                                                                ....:.,+.J                                                        .

4 .. .. . . . . . . . . . . . .,, . _ . . , .

;                                                    .r t. c . . . . . . . . .                      ., , . . .u . . : . , . . . . v .

s.

                                                                                                                                                                                                                                  ,___e._u-l                                                   The Reference 3 4 analyses demonstrate that eith one RCS relief O f 6'a'l 2                                                    valve or the depressurized RCS and RCS vent can Isaintain RCS l                                                  pressure 6+*.- i...       . -

below limits whens,My is are actuated

                                                                                                                                                                                                            ;.x lTI m erih gg,y,N8 the LCO allows only one l                                                  lTI m ;.r4 ne m.m - -                                                                        -- ----
                                                                                                                                                                           - -- -- PERABLE during the
 }                                                                                                        (indli-/%,doNeA/00 i

(continued) MARK UP 0F NUREG 1431 BASES B 3.4 68 5/15/97

M B 3.4.12 BASES APPLICABLE Heat Inout Tyne Transients (continued) SAFETY ANALYSES L40P M H0 DES. Since neither one RCS relief valve nor the RCS vent can handle the pressure transient M re;d fre accumulator injection when RCS temperaNre is low, the LCO also requires the accumulators- isolation when accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed in the PTLR. Se isolated accumulators must have their discharge valves closed and the valve power supply breakers fixed in their open positions. Oi rely;;; ;t.e tt.; cffat of ;G;;;1;ter di;;terg i; :nr e Terr..;r "00 t-: ;-;r;tur; r;n;; (170"I a.d b;1;; then . trat ;f tra LOO (070"i s.d bel s). Fracture mechanics analyses established the temperature of L40P M Applicability at EM M Oc cen;;;;ree; ef ; ;;.;11 bisk le;; cf ;eela.t e;;id,.rt (LOCA) in L'"" ".^.,^0

                                 . 4 ca.fe ; t; 10 0l" 00.40 x: 10 0~, 00. Apgr. dix X

(";f;. O sd 0), ru,uir ..t; by P.nia; ; =xi;= cf ex ll"I p;;p

                     ; .d we ;tergir.; p-- 0"C"J"LC .ad 0! ntatic cati;d.
                                                                ,, w           I-         & 2AGovl On;  u PORY Performance                                r   r ""

rdliney The fracture mechanics anal) es show that the vess is protected when the PORVs are set to open at or below the lim t3 shown in the PTLR. The setpoints are derived by analyses pat modeg performance of the . - L" Sy;ts - li;'tir, m.

                         -           - transient ofih, assuming           g          and        04-3,4-o03 444 A/C/0injecting into the RCS a*         These analyses consider pressure overshoot and undershoot beyond the PORV opening and closing, resulting from signal processing and valve stroke times. The PORV setpoints at or below the derived limit ensures the Reference 1 P/T limits will be set.

The PORV setpoints in the PTLR will be updated when the revised P/T limits conflict with the L40P v analysis limits. Tne P/T a limits are periodically modified as the reactor vessel material (continued) MARK UP OF NUREG 1431 BASES B 3.4 69 5/15/97

a. s .: wy e.w

, B 3.4.12 a rec lline) t BASES (continued) v 4 SURVEILLANCE SR 3.4.12.1. /SR 3.4.12.2)' and SR 3.4.12.3 d8MSe/ REQUIREMENTS a To minimize the potential for a low temperature overpressure event by limiting the mass input capability, a maximum of en : N

                                                                             . . ~ . . . . .

m 9. . . pump +- . 1 - . .: .- are verified . incapable of injecting into the

RCS and the accumulator discharge isolation valves are verified
closed '<

                                                                                              ;r.d 1;;Md r,;t.                                                   -               .
                                                                                                                                                                                                -        ~      .                 -

1 (rM)

~~

Tht. HPI pump charging cj).g,4.gg

; .                                                                                                                                                                                               re rendered incapable O s.+. n-1

of 1 jecting into the RCS through removing the power from the pumps by r6cking the breakers out under administrative control.

                                                                                                  ..t,             e                   ,
                                                                                                                                                                                                                             .            e ,

i

                                                                             .                           e    n s                  .                                                                      <

a

k. elterret; atkd ;f L"" centiel => k e ,,1;y;d usirs et 1;;st t.e ind;p;ndent n;n t; pr;;;nt ; pg ; tert ;;;h ,

thet ; ; irs,i; f;ilur; er ;irsi; ntien will r.r,t r;; ult in enu. s..,~.,.,.

                                                                               .             . . . . . .      .u.    .~ ..n. s~,,.

w. _u. .__,,ca

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

c__..a ps;p centr 1 ; witch kirs pieced in pull to 1;d end et 1;nt , ex niv; ir, tk di;c;r, fi;w p;th kirs cised. ! 4 The Frequency of 12 hours is sufficient. considering other indications and alarms t...6,..,mt,,...,,..... . available to the operator in the control room, to verify the required status of the equipment. m m..+. i Each required RHR suction relief valve shall be demonstrated OPERABLE by verifying its "J:", ;;; tie. v;1v; end RfR suction 1 (continued) 4 MARK UP 0F NUREG 1431 BASES B 3.4 77 5/15/97

                                                                                                                                                                                                 .. c to E WE =#J d bEspI s

B 3.4.12 BASES REFERENCES 8. ASME, Boiler and Pressure Vessel Code, Section XI. (continued) g '

                                                                                     ^ - ;> :.:;+O e -w rw 0 ;-       '? " . ' ; g f r, i .,', r - ,yg 2 M             . s .~ .M-3 7 :p :r.:.
                                                                                                                                                                                $ 6  &                      *
                                                                                                                              $                                                                                 a Dc ne n M=n AM &at 2, nn.

2 N 4 MARK UP 0F NUREG 1431 BASES B 3.4 81 5/15/97 4

                                                                               -e- ~

1 l l ADDITIONAL INFORMATION COVER SHEET l ADDITIONAL INFORMATION NO: CA 3.4-004 APPLICABILITY: CA, CP, DC, WC l REQUEST: This item covers the following changes:

1. Revise ITS 3.4.14 (and corresponding CTS mark-ups) to reflect that the RHR s'iction isolation valves from the RCS are remote-manual, not automatic (Not applicable to DCPP, WCGS, and CPSES).

2. Define an OPERABLE RCP in ITS 3.4.4 LCO Bases as defined in ITS 3.4.5 and 3.4.6 LCO Bases (Not applicable to DCPP).

3. Revise the ITS 3.4.9 Bases for Required Action A.3 to match the Bases changes made for ITS 3.4.5 Required Action C.2 per approved traveler TSTF-87 with regard to ways to make the Rod Control System incapable of rod withdrawal. l

4. Revise the Applicability Bases for ITS 3.4.10 to reflect the change to the Note in Enclosure SA under JFD 3.4-18, i.e., the Note allows entry into MODE 3. MODE 4 i should be struck-through as it was in Enclosure SA since the pressurizer safety LCO is not applicable in MODE 4 (Not applicable to DCPP, WCGS, and CPSES). ;

ATTACHED PAGES: Attachment 10, CTS 3/4.4 - ITS 3.4 Enclosure 2, page 3/4 4-19 INSERTS A and D Enclosure SA, page 3.4-33 Enclosure SB, pages B 3.4-19, B 3.4-46, B 3.4-50, and B 3.4-93 I l e

INSERT A FOR PAGE 3/4 4-19 y-~s isolate the high pressure portion of the affected system from the low pressure 4-//-LS portion within 4 hours by use of at least one deactivate =:=.;;i- 6r check c5Ef-ood valve ## and within 72 hours reduce the leakage rate t within lim ~ s; 4-/.2-M otherwise re~de manul INSERT B FOR PAGE 3/4 4-19 BB-PV-8702A/B and EJ-HV-8701 A/B ar6 excluded in MODE 4 when in, or /-0/-LJ during the transition to or from, the RHR mode of operation. s INSERT C FOR PAGE 3/4 4-19 Each valve used to satisfy this ACTION must have been verified to meet l-/3-Al Surveillance Requirement 4.4.6.2.2 and be in the reactor coolant pressure boundary. INSERT D FOR PAGE 3/4 4-19 With the RHR suction isolation valve interlock function inoperable, isolate f-22-/M the affected penetration by use of one deactivatedComsticjalve witnm 4 cg_7,4_og4 hours. re d u nu j O

l RCS PIV Leakage 1 3.4.14 ACTION' (continued) ' COWITION REQUIRED ACTION COMPLETION TIME l A. (continued) A.1 Isolate the high pressure 4 hours' portion of the affected system from the low pressure portion by use of one e ksed rPS$ ur.=1. deactivated r 4 p-anual cA-9,po4

                      .                      -eutematie or check valve.

M A.2 Ini;t; tre Pi@ i,7;;;;r; 72 t.;;r; sB PSs p; tim of tt.; effat;d

                                               ;y;ti 776 tt.; ici pruar;

. prtin b ax ;f ; ;nad

                                               ;i ned 27.21. ds;tivated sta;;;. ;r ; tad v;1;;.

IE p .

                                                                                                    ;Bt V

B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time for Condition A not M inet. B.2 Be in MODE 5. 36 hours 3 RlR - - - C.1 Isolate the affected 4 hours m B PSt M Sy;ts atulence penetration by use of one interlock function ci;;;d s7.21 er deactivated e3.4 11N inoperable. -automa64ex valve. ree ~nu. l cg_y,4,4 MARK UP OF WOG STS REV 1 (NUREG 1431) 3.4 33 5/15/97

-n

   &:.n O ,21         d eu . w;.~,, L aw rh ..w :G1.4.9 16 ar.w$ .:   p .y; x- 2ia :du 34           6. ,_ #4 # 6A r-s.w.,- .w .--     n,  .:. .
                                                                                                                                                . l
     ~.                .
                                                                                                                     ~        m       -~
                                                                                                       "    'RCS Loops    MODES 1 and 2 B 3.4.4           i BASES APPLICABLE SAFETY ANALYSES                               -

(continued) (~ $ S.f. 6% / The plant is designed to operate with all RCS loops in operation to maintain DISR above theM during all normal operations and anticipated transients. By ensuring heat transfer in the nucleate boiling region, adequate heat transfer is provided between the fuel cladding and the reactor coolant. RCS Loops - H0 DES 1 and 2 satisfy Criterion 2 of tt.; %"C .";1 icy N LCO - The purpose of this LCO is to require an adequate forced flow I

i. f rate for core heat removal. Flow is ra 'sented by the number of RCPs in operation for removal of heat L, the SGs. To meet safety

                                           .                  analysis acceptance criteria for DNB,              pumps are required at
                                                       -      rated power.

An OPER,9tE RCS loop consists of an OPERABLE RCP in ep;reti;n presiding farced 'h; fer t.;;t tren; pert and an OPERABLE SG in cordarse with the Steam Gipnerator Tube Sury Progras akn /cr i;r ofE' gAfLE if T+ T.r cojajle oh111arr he fewere)ed it able le frevr/e &cea ffw. f CA-2.+-ocq APPLICABILITY, In M00ES.1 and 2, the reactorg 4e critica15 e d the; has the potential to produce maximum TERMAL POWER. Thus, to ensure that I the assumptions 'of the accident analyses remain va111 all RCS l loops are required to be OPERABLE and in operation in these MODES to prevent DNB and core damage. l The decay heat production rate is much lower than the full power i heat rate. As such, the forced circulation flow and heat sink requirements are reduced for lower, noncritical MODES as indicated by the LCOs for MODES 3, 4, and 5. Operation in other H00ES is covered by: LCO 3.4.5, "RCS Loops H0DE 3": ! LCO 3.4.6, "RCS Loops H00E 4*: LCO 3.4.7, "RCS Loops MODE 5. Loops Filled"; LCO 3.4.8,_"RCS Loops MODE 5, Loops Not Filled": LCO 3.9.5, " Residual Heat Removal (RIR) and Coolant l Circulation-High Water Level" (MODE 6); and l (continued) t MARK UP OF NUREG 1431 BASES B 3.4 19 5/15/97

Pressurizer B 3.4.9 BASES AvTIONS A.15 end A.2m (continued) unit must be brought to H0DE 3, with m

                              ,,.a.a.m         c,    . . , . . .y y the r;;;ter trip hr;de ; ep;n, .;ithin 0 heur; end to MODE 4 gg witMn 12 hours. This takes the unit out of the applicable F8Nh8        HOQ .r.d r;;ter;; tt~ r" ^; ;r "-- 4 thin th; t;;nd; cf O'8dden-/

I th; ;;f;ty :nely;;;. (e.g 1 7de-ene rz - all cKbMg &y The allowed Completion e ni 3 %e RT8.r -e is.r3 +M

                                                                           'are#e"a'               e           _

perating experience, to reach the requir conditions froiiiTun

power conditions in an orderly manner and without challenging plant systems. IL1 If one required group of M pressurizer heaters is 1 inoperable, restoration is required within 72 hours. The Completion Time of 72 heurs is reasonable considering the anticipation that a demand caused by loss of offsite power would be unlikely in this period. Pressure control may be maintained duri ng thi s time using .~. - . ... . . ,~ .. . ..,...m

. E

c. .

                                     ..m   -

c..n w.., ., 4 .. .n., . ., . .. i . - . C.1 and C.2 If one group of M pressurizer heaters are inoperable and I cannot be restored in the allowed Completion Time of Required Action B.1, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours and to H00E 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.4.9.1 REQUIREMENTS This SR requires that during steady state operation, pressurizer level is maintained below the nominal upper limit to provide a (continued) MARK UP OF NUREG 1431 BASES B 3.4 46 5/15/97

Pressurizer Safety Valves B 3.4.10 BASES APPLICABLE Pressurizer safety valves satisfy Criterion 3 of tM OC Pa14ey SAFETY ANALYSES Statement- sammmmmmmmmmmmmmmmme , (continued) LCO The . pressurizer safety valves are set to open at the RCS design pressure (500 pi; w e ), and within the ASE specified tolerance, to avoid exceeding. the maximuis design pressure SL, to maintsin accident analyses assumptions, and to comply with ASE requirements. The upper and lower pressure tolerance limits are based on the i 12 tolerance requirements (Ref.1) for lifting pressures above 1000 psig. The limit protected by this Specification is the reactor coolant pressure boundary (RCPB) SL of 110% of design pressure. Inoperability of one or more valves could . result in exceeding the SL if a transient were to occur. The consequences of exceeding the ASE pressure limit could include damage to one or more RCS components, increased leakage, or additional stress analysis being required prior to resumption of reactor operation. APPLICABILITY In MCOES 1, 2, and 3 ;ad prtiea; cf "000 4 ;Lv; tM L~ ; ;ia; t; gr;ture, OPERABILITY of M valves is required because the combined capacity is required to keep reactor coolant pressure below 110% of its design value during certain accidents. MODE 3

                                                            .ad prtiea; cf "00C 4 er; 3 conservatively included, although the listed accidents may not require the safety valves for protection.                                                             .

The m-_m___ LCO is not applicable in H0DE 4 WMa ell Z ;;1d 1;; f) 7.4,6e-/

                                                                                                            - '- H00E 5         4 -

2 - ,... .. because ETOP i . is previd;d. Overpressure protection is not required in H0DE 6 with g reactor vessel head detensioned. V The Note allows entry into H00Ef 3.#$(with the lift settings CA-3'.4-044 outside the LCO limits. This permits testing and examination of the safety valves at high pressure and tesperature near their normal operating range, but only after the valves have had a preliminary cold setting. The cold setting gives assurance that the valves are OPERABLE near their design condition. Only one (continued) MARK UP OF NUREG 1431 BASES B 3.4 50 5/15/97

RCS PIV Leakage B 3.4.14 I ACTIONS B.1 and B.2 (continued) also reduces the potential for a LOCA outside the containment. The allowed Completion Times are reasonable based on operating

experience, to reach the sequired plant conditions from full power conditions in an orderly manner and without challenging plant systems.

i C.l i j The inoperability of the RlR esteeksure - interlock re.-i. .; tre ~" n; tim ini; tim v;1va tra;per,1; cf i ini;tir.ii 17. rupan't; ; hi7 presar; G.-Jitia ;d pr;a..tir.;;

i. ~ .

inadvertent opening of the valves at RCS pressures in j excess of the RIR systemt design pressure. If the RIR

euteeksure- ' -

interlock is inoperable, i operation may continue as long as the affected RlR suction penetration is closed by at least one eksed-1aanuel-or ! deactivated . automatic valve within 4 hours. This Action CA'7 +-def-accomplishes the pu se of the - wt al; ar; furati m . l

kremrle mamaj 1 SURVEILLANCE SR 3.4.14.1 REQUIREMENTS Performance of leakage testing on each RCS PIV;-9,.G.J.p SXf./6./

used to satisfy Required Action A.1 ;.-J Rc,. ired 'etim ".2 is required to verify that leakage is below the specified limit and to identify each leaking valve. The leakage limit of 0.5 gpa per inch of nominal valve diameter up to 5 gpa maximum applies to each valve. Leakage testing requires a stable pressure condition. For the two PIVs in series, the leakage requirement applies to each valve individually and not to the combined leakage across both valves. If the PIVs are not individually leakage tested, one valve may have failed completely and not be detected if the other valve in series meets the leakage requirement. In this situation, the protection provided by redundant valves would be lost. Testing is to be perfonned every 3 months, a typical refueling cycle, if the plant does not go into MODE 5 for at least 7 days. (continued) > NARK UP OF NUREG 1431 BASES B 3.4 93 5/15/97

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO. TR 3.4-004, TR 3.4-005, APPLICABILITY: TR 3.4-006, TR 3.4-009 CA, CP, DC, WC REQUEST: Revise the Traveler Status Sheet to reflect that TSTF-54 Rev.1, TSTF-87 Rev. 2, TSTF-136, TSTF-137, TSTF-153, TSTF-162, and TSTF-233 (was WOG-67) are approved by NRC. Add Rev.1 to TSTF-94 and TSTF-151. ATTACHED PAGES: 0, CTS 3/4.4 - ITS 3.4 B, page 4 Enclosure SA, Traveler Status Sheet A, page 5 8, page 4 l 4 i i

                                                                                                                                                                                                          -i     1
                                                                                                                                                                                                                 )

CONVERSION COMPARISON TABLE - CURRENT TS 3/4.4 Page 4 of 13 TECH SPEC CHANGE APPLICABILITY NUE ER DESCRIPTION DIABLO CANYON CONANCHE PEAK WOLF CREEK CALLAWAY 3-01 This change revises ACTION from the " pressurizer otherwise Yes Yes Yes Yes LS-4 inoperable" to " water level not within limit". 3-02 The method of verifying heater capacity is moved to the Yes Yes Yes Yes LG Bases. 3-03 This change extends the reysurizer heaters surveillance Yes Yes Yes No - Already part LS 28 interval from 92 days to 18 months. g c.g_L--gf of current TS per Amenchnent 105. 3 04 The ACTION is modified to reflect generic wording to assure Yes Yes Yes Yes LS-29 that the rods are fully inserted and cannot be withdrawn. This change is consistent with traveler TSTF-87 L. 1. -fg-2gg 3-05 This change adds a requirement to current TS 3.4.3 LCO. The Yes No - Pressurizer No - Pressurizer No - Pressurizer M change requires the pressurizer heaters to be capable of heaters are heaters are heaters are being powered from an emergency power source. normally aligned to normally aligned to normally aligned to emergency power. emergency power. emergency power. 4 01 An allowance has been added that separate condition entry is Yes Yes Yes Yes LS-5 allowed for each Power Operated Relief Valve (PORV). 4 02 The allowance for PORY inoperability has been exparded from Yes Yes Yes Yes LS-6 previously only allowing block valve isolation when the PORV was inoperable for seat leakage to allowing block valve isolation when the PORY is still capable of manually being cycled. 4-03 The times to reduce MODE are decreased by one hour. Yes Yes Yes Yes M 4 04 This change moves the requirement to perform channel Yes - Moved to Yes - Moved to TRM. Yes - Moved to Yes - Moved to LG calibration of PORV actuation instrumentation to a licensee FSAR. USAR. FSAR. controlled document. CONVERSION COMPARISON TABLE - CURRENT TS 5/15/97

l

INDUSTRY TRAVELERS' APPLICABLE TO SECTION 3.4 { TRAVFlFR # STATUS DIFFERENCE # COMENTS

TSTF 26 Incorporated 3.4 32 Approved by NRC.

TSTF 27. Rev. N Incorporated 3.4 33 Ajprwell y A/4d. -((f jg TSTF 28 Incorporated 3.4 22 Approved by NRC. TSTF 54 Rev. 1 Incorporated NA Ns~M3/ chin #EEnly. g rg s4g

                                                                                                                                  .i TSTF 60              Incorporated                  3.4 15             Approved by NRC.                                       O TSTF 61              Not Incorporated                 NA              Minor change that is adequately addressed in the Bases.

TSTF 87. Rev. ncorporated 3.4 31 A//rWM yl A/AO - 7g-3.Q f Incorporated 3.4 17 N M $r"o Ned Ior Y ia'way in OL TSTF93.Rev.$ L3 & 2.+,9-3 Amendment No.105. g2.9 : 3- - TSTF94;/ev,i Not Incorporated NA Retained current TS. W-2.4( h

2:' !?E,$= l Ixep d.d 5.4 ^^. g gfl-l -re n - r TSTF 108, Rev. 1 Not Incorporated NA LCO 3.4.19 does not apply. M TSTF113.Rev.[3corporated d ;T,4,/4-/[$.2f;//-3

3.4 39 -'Z' 2. ' 3//- TSTF 114 Incorporated NA Approved by NRC: Bases 3.4.7 changes only. u TSTF-116, Rev. [ brporated 3.4 36 j9;T4/F-2 -en a mU TSTF 136 Incorporated NA B E Y E N N .I changes on1y. , _ 7g-y4e .n TSTF1b7 Incorporated NA B@sTUNchfrheNnly. -rg g,4 g t TSTF 138 Not Incorporated NA Inconsistent with RCS loops requirements of ITS 3.4.5 and ( 3.4.6. I TSTF 151gev. / Incorporated NA Bases 3.4.11 changes only. TR.7.hP)f fl TSTF 153 Incorporated 3.4 01 AN '* V,* I Ir A/AC- 74-3,W TSTF 162 y Incorporated NA Es 3N.9/chInN>'only.7AL3+g. TJYF-REV / N .+, / 2 .~1 *1 sm 1

                                                                                                                             -j Eq             Incorporated          2 4 4 0 . 3.4 45,         See also CRs .41B and 3."4f20.~'
    '7%Tg-air
     - ..                  Incorporated                 3.4 35                         g y,4ff-p                --co f.f gjj
     ,2P7,F-asy                                                                                                                  r

9. - . . . . r! .- Incorporated 3.4 10 DCPP only./)pr,ve/ lyn #d.ry_S.4 q 4

WOG 87, ge v. 2. Incorporated 3.4 47 $ 7,f:// 7,f T.fm i

l HARK UP OF WOG STS REV 1 (NUREG 1431) 5/15/97 i

l. .

CHANGE Nl#EER JUSTIFICATION bvH 3.4 30 An LCO 3.0.4 exception is added to the Actions for LC0 3.4.12. This is consistent with the current licensing basis. Additionally, circumstances could arise where increasing MODE would reduce the risk of a low temperature overpressurization event. In these cases it would i be unwise to maintain the plant in a lower MODE configuration. Increasing plant MODE may also be the expedient way to exit a low i temperature overpressurization potential when operating within a Condition. This option should be retained as exists in the current Technical Specifications. l 3.4 31 These ACTIONS in ITS 3.4.5 and 3.4.9 are modified to reflect their LCO. ) j The position of the reactor trip breakers and the power supply status ! i of the CRDHs are not LCO requirements; therefore, the CONDITIONS and ACTIONS are revised. As worded in NUREG 1431 Rev.1. these ACTIONS ) could preclude certain testing in H0DE 3. A more generic action, which assures rods can not be withdrawn, replaces the specific method of j precluding rod withdrawal. The specific methods are added as examples l to the Bases. The revised ACTIONS still assure rod withdrawal is precluded and this detail is not required to be in the.TS to provide l adequate protection of the public health and safety. No technical changes result from this change. These changes are consistent with traveler TSTF 87,O;.1. ~72-3'f-@yt-3.4 32 In accordance with industry traveler TSTF 26, the ACTION would be changed to specify taking the plant to a H0DE for which the LCO is not applicable. This change maintains the consistency between the Mode of Applicability and the Required Action which requires the Mode of Applicability to be exited. 3.4 33 The Frequency of SR 3.4.2.1 to verify operating RCS loop average temperature at or above [551]*F is changed to 12 hours from the current surveillance frequency of 30 minutes. . The SR to verify operating loop average temperatures every 12 hours @ -"##WP , ". :;;:0 t: ; r;.c.; ---

                                                              =ti:t:t ditt; s' tle :.C^ ;;d considers indications and alarms that are continuously available to the operator in the control room. O E This change is based on industry traveler TSTF 27, to. 2.         O 7.I 4   /4 3.4 34                          Not applicable to Callaway. See Conversion Comparison Table (Enclosure 68).

3.4 35 This change adds a Note to SR 3.4.11.1 and SR 3.4.11.2 stating that the SRs are only required to be performed in H0 DES 1 and 2. The Actions Note, "LCO 3.0.4 is not applicable," is intended to allow H0DE changes for testing purposes (per the Bases). This allowance is properly presented as an SR Note. A properly placed exception (i.e., an SR Note exception) would not allow the SR to be considered to be met until the appropriate conditions were available for it to be perfomed without entering the Actions. The Note to these SRs would allow startup in l H0DE 3 if the SR had not been performed during the required frequency. I JUSTIFICATION FOR DIFFERENCES - TS 5 5/15/97

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

                         .      y                                                                                                                                                                                      v3    ,

CONVERSION COPFARISON TABLE FOR DIFFERENCES FRON MREG-1431 Page 4 of 7 ' SECTION 3.4 ! l r DIFFERENCE FRON NUREG-1431 APPLICABILITY NUPSER DESCRIPTION DIABLO CANYON CONANCE PEAK WOLF CREEK CALLAIMY 3.4-27 Applicability for LCO 3.4.12 MODE 6 is revised to include Yes No No No an additional qualification if the head closure bolts are not fully de-tensioned per DCPP CTS. , 3.4-28 This change adds a DCPP-specific description of a secured Yes No No ~ No l open valve. 3.4-29 The use of Channel Functional Test (CFT) would be retained Yes No No No f from the current DCPP TS to the improved TS. l 3.4-30 An LCO 3.0.4 exception is added to the Actions of No - This change is Yes Yes Yes ! LCO 3.4.12. out of scope for l DCPP. j 3.4-31 Condition C and Required Action D.1 of ITS 3.4.5 and . Yes Yes Yes Yes } Condition A of ITS 3.4.9 are modified to reflect generic , wording to assure that the rods are fully inserted and cannot be withdrawn. These changes are consistent with r traveler TSTF-87. 6 . i. ~T/ d N Q f 3.4-32 In accordance with industry traveler TSTF-26, the ACTION Yes Yes Yes Yes would be changed to specify taking the plant to a MODE for l which the LCO is not applicable. i l 3.4-33 The Frequency of SR 3.4.2.1 is changed to "12 hours." Yes Yes Yes Yes -! This change is based on industry traveler TSTF-27. ! t 4 3.4-34 Retains CPSES current TS which requires that the precision No Yes No No ) RCS flow measurement be performed prior to exceeding 85% RTP. 3.4-35 Adds a Note to SR 3.'4.11.1 and SR 3.4.11.2 stating that Yes Yes Yes Yes the SRs are only required to be performed in MODES 1 and 2. t 3.4-36 SR 3.4.13.1 and ACTIONS for LCO 3.4.15 are revised with Yes Yes Yes Yes the addition of a Note per traveler TSTF-116. , CONVERSION CONPARISON TABLE - NUREG-1431 5/15/97 !

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: WC 3.4-002 APPLICABILITY: CA, CP, DC, WC REQUEST: Clarify ITS 3.4.9 Applicability Bases to state the pressurizer heaters are capable of I being powered from either the offsite power source or the emergency power supply. ATTACHED PAGES: Attachment 10, CTS 3/4.4 - ITS 3.4 Enclosure SB, page B 3.4-45

Pressurizer B 3.4.9 l x BASES LCO the RCS near normal operating pressure when accounting for heat I (continued) losses through the pressurizer insulation. By maintaining the ' pressure near the operating conditions, a wide margin to subcooling can be obtained in the loops. Oc en ct i ;ig, nia; cf 125 kW i; deri;;d free th; ;;; ef ;;nn tieter; r;ted et 17.^ kW ;;;h. Oc 1---;nt racid te nint;in pr;;;;r; i; i gr.ii,; er, tra P;;t 1;;;;;. , 1 APPLICABILITY The need for pressure control is most pertinent when core heat can cause the greatest effect on RCS temperature, resulting in the greatest effect on pressurizer level and RCS pressure control. Thus, applicability has been designated for MODES 1 and 2. The applicability is also provided for MODE 3. The purpose is to prevent solid water RCS operation during heatup and cooldown to avoid rapid pressure rises caused by normal operational perturbation, such as reactor coolant pump startup. f e(f44,-4As effef4e fewar rama or Or- WOWG In MODES 1, 2 and 3 there is 33 need to maintain the avail ability of pressurizer heaters, capable of being powered f fronVen emergency power supply. In the event of a loss of offsite power, the initial conditions of these MODES give the greatest demand for maintaining the RCS in a hot pressurized condition with loop subcooling for an extended period. For MODE 4, 5 or 6, it is not necessary to control pressure (by heaters) to ensure loop subcooling for heat transfer when the Residual Heat Removal (RHR) System is in service, and therefore, the LCO is not applicable. ACTIONS A.17 end A.21!EEWEEEEEE Pressurizer water level control malfunctions or other plant evolutions may result in a pressurizer water level above the nominal upper limit, even with the plant at steady state conditions. Normally the plant will trip in this event since the upper limit of this LCO is the same as the Pressurizer Water Level High Trip. If the pressurizer water level is not within the limit, action must be taken to :. eta . 4 :..< + a ,e'aa,~me C.*S.& M'WP r;;ter; tre plent te egretien within tre bar.t of tra ;;f;ty ;raly;;;. To achieve this status, w ..ir 0,% e the

~,

(continued) MARK UP OF NUREG 1431 BASES B 3.4 45 5/15/97

ADDITIONAL INFORMATION COVER SHEET l l ADDITIONAL INFORMATION NO: WC 3.4-007 APPLICABILITY: CA, DC, WC REQUEST: Revise the Frequency of ITS SR 3.4.11.2 to read: "In accordance with the Inservice Testing Program" consistent with the CTS. ATTACHED PAGES: Attachment 10, CTS 3/4.4 - ITS 3.4 Enclosure 5A, page 3.4-23 Enclosure 58, page B 3.4-61 Enclosure 6A, page 8 Enclosure 68, page 7 l 1 1 l

Pressurizer PORVs

      ~ .                                                                                                                                                           3.4.11 r\
    \             SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                                                      FREQUENCY SR 3.4.H.1                                  -

NOTET - - - -- 3 Not required to be met 2 5tUffEl with M3~4 47' block valve closed in accordance with the Q 5.+.// f-Required Actiogf EsMiti;- 0 ;r C. -/S T.r / Co. IU#O g .m . .;.:. m .a.e . . . , . . .r-p a.n. . g. ,. , c , , a.u n3.4 35? l

                                                    .w -

i Perform a cosplete cycle of each block _ 92 days valve,- - m.--- - - - - , - -- - E3;4 471

                                                                                                                                                      &:f.4.// 4^ -C' l %"-
                                                                                                                                              % accorbee 3W                                     ,

SR 3.4.11.2 "^YA S* ~ l 4 - .

                                                        ...__.---.._..._-_.._._..s.....-m.

L.rewice n3.4 35 e E . . . ~ .. . . . . . . , . ,. . . , , . .

                                                                                                                                                'fe.rH} fn}m ..

Perform a complete cycle of each PORV. "J nath; -

                                                                                                                                                 ,                                ?BP      >

Z 3.4.11.0 "cifer; e cWictc cyci; ef ;;;h ;;l;mid eir 10 7enths nab PS: ' centr;l velv; erd check velv; en the air ecca;;.uleters ir "TV centrol syst;;;. Z 3.4.11.4 VCiify "0"S; ord bleCk velvC3 Gr; Cep0hl; Cf 10 ser.th- xB PS bein POWCred fie; saiyCn;j p0W;r 30uiC;3.

    ,r x MARK UP OF WOG STS REV 1 (NUREG 1431)                              3.4 23                                                                        5/15/97

Pressurizer PORVs B 3.4.11 c,, BASES (continued) SURVEILLANCE SL 3.4.11.1 REQUIREMENTS Block ralve cycling verifies that the valve (s) can be

                                                                      ,,          ___s              ,

c . .. .w.

                                                                                                                                                                                                                                                             ,a    .n_

m .,m as r Frequency of 92 days is the ASE Code, on XI O M // 4

                                                                                           ,,       .u        u,,            p    o,i. u. 4, w.____              ,,.. a ._ 2.._,.._1.-                                  -
                                                                                                                                                                                                                           ,,,,,u,
                                                                                                                                                                                                               . .                      ...b.w.

(Ref. 3 E,.). , .. . . _ - _ . _._ . . . , , . . _ . _ , - > m- - - . - - . - - - - m. 4.*-r--'" . ... 6 "'

                                                                                        , , u4
                                                                                              "W           "'"'J                          *J * ' W8              6' N

3'#V#'I"A I 3 U3 b' N

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y .7 ww w ww.s .'."_"""l7 . U%151 @ b.5.^

                                                                                                                                                                                                     - ' - - - S a .m
                                                                                                                                                                                             .' 9. Ubl4        y as y r,.

Jm vw ____....,s. _ _ _; s sL. nanu A_ a

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L._ w wwww wwww g wv gr i-,w wi wisi = weis bW W .... I ws - -ww s ww w . w . wy wg wu gugu a p_.. jag ga AL_ L1__L .._1,,_ Td-m --_A-__ AL_ hanu J ____ _ vww, ww w a , www ww e is. 4F m ,o wyWh p wv,W WW 5%5 _7_.I b5Ub W 5 WhW4 V455Yh J..,,7 "fD L ..m m - 13Ab 4e ml I k_.d & bd n Ob .3 S -- - ' S a 3dm4ha # M 08 % 1- ! yw , . _ . wy WFly e u s vw nwa wew wieas w = -w wu.-.--.-sw u e av v we g 5 e,85 g bW m'J'h aapud &km _..__.._ w..w .L1_ L- . . u. w. .w

                                                                                        .                               f*_--_--.,.

i. m am. ., - ,a, d, n,.#1.

                                                                                                                                                                    .        .J.w,..
                                                                                                                                                                                   . s .m f"   . A L -
                                                                                                                                                                                                     . . . . _ . . , . . . .          A.L--.                           1 I

m_

                                           ...___m.._.... . w
                                                                         ,1ae.,4
                                                                                                          .- .2              L ._
                                                                                                                                                 -_,_i_,

L. m- ______,__ ,

v. .

                                           ,_---.*.1.,u, h1.,_u.L_ . . .w.v.                     - - ,vu- . ..'6vi u'b e up                                           vu        vs     ".v.- ^^^"s run           bw m mm.m a .m s s=                _.          1             s3     _                         ,_AJ__                    _*     AL_             sa _ _ . . 3__    J    a_ii___

vruv, u. ... .. si.... ,,....,,, v. .. nw . . ns. . .. - ,

.:i"a u 5iu. p4: Gen-/ l j

tar.cSR - hNote 3 modifies this SR by stating that s not required thM.!H2 g g,.g h h,h* he equired Acti - .r. l,g

                                                                                                                                                                                                                                                       & s.4,ll-+

1 . . . )

e n,- -I el 4 toe ydye in Q24//-+ rs c i-/Gn inc euwr Oe l SR 3.4.11.2 Ffrk an uni.rofal,le feaj. fm , -1),e Ac.r.tinee +Ae foRV fr er/ redy SR 3.4.11.2 requires a complete cyc "N- Operating a i PORV through one complete cycle ensures that_the__PORV can be _

' - _~ aanually ac%ted for mitigation of an SGTR. ": F 07. - y e: h/c-Jf-ooy

                                           = r=th:                                           .

4 cyci' r.d ir.dustry _,. +.,a 1:. 'eeed +4-- cr. Ofenhryex: ti c ' c".clir.;;har

                                                                                                                                                           <   n e   ne    e         s    Aous., 4Ad 4-here 4lver usun         a.r.r 4Ae'3 TIN ~II 5 Dlen *rform e                                                                                             *+      Eh*                          utred rme <ce
                                               -rta                  re uane                                                               Jia %a eo                                              1:4 y fop  y         regen                                                                                                                                                                                                               .
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        /

2 .

                                              .a.            .
                                                                                      . ,.                 . . . : .                                - i. a,,.         . .               ....n..          .is.      .
                                                                                                                                                                                                                          ,...,,p
                                              .. .. . . . .                                - ..,            . -                               .        . . .. - .                                  .. . .             .m.       . ~,.

j w.4 . 3 t. f , . c s . .. n. ... s.,g,c.s ..o a os u. , . . (.2, .;

.                                                                                            A

(

,                                                                                                                                                                                                                          (continued) i i

MARK UP OF NUREG 1431 BASES B 3.4 61 5/15/97 1

CHANGE NUMBER JUSTIFICATION i' ~ 3.4 49 LCO 3.4.12. "[COMS]", provides four different methods for pressure relief. Any of the four methods may be used. However, Surveillance Requirement 3,4.12.5 requires testirig whether or not the equipment is g, being credited to meet the LCO. The proposed change adds the word 3 " required" to the Surveillance to exempt its performance if the equipment to be tested is not being used to meet the LCO. In addition, two editorial changes were made. The LCO requirement presentation was clarified. Also, the Note to SR 3.4.12.8 was revised to replace

                    " required to be met" with " required to be performed" since the
                    " performed" nomenclature is appropriate here, consistent with the CTS.

This change is consistent with traveler ':E 100. / 57F-:af0, f)?A/2-/ 3.4 50 This change is consistent with current TS SR 4.4.9.3.3. The 12 hour l frequency applies to vent pathway (s) that are not locked, sealed, or l otherwise secured in the open position. The wording added to ITS l SR 3.4.12.5 is also consistent with the format used in similar ITS 3.6 SRs. The 31 day frequency is also revised to be consistent with current TS SR 4.4.9.3.3. 3.4 51' Not applicable to Callaway. See Conversion Comparison Table (Enclosure 68). 3.4 52' Not applicable to Callaway. See Conversion Comparison Table (Enclosure 6B).

9. +-ss .ruseer 2.+-ss N&~3 447 S?+--s+ Afd ag/ rude de Ody. See Onvenion 0,,ng,u,,

b c-2. 4-a3 j 7hla (enc lmaraif). l JUSTIFICATION FOR DIFFERENCES - TS 8 5/15/97

l i-i. l ' INSERT 3.4-53 Wc-2.+-n y o This change revises the ITS SR 3.4.11.2 Frequency from "18 months" to "In accordance with the Inservice Testing Program." The CTS for this surveillance establishes the frequency as l being per the IST Program [ CTS SR 4.4.4.1). t I ( i f l

CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431 Page 7 of 7 SECTION 3.4 DIFFERENCE FROM NUREG-1431 APPLICABILITY NIM ER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY r DY~ bc- ALL-MS No 3.4-51 The Note for SR 3.4.1.4 is remov x!. This is consistent Yes No No with DCPP CTS 4.2.3.5. DCPP cor ts a measured RCS total flow rate verification on an@ th frequency. , , , ,_ 3.4-52 Consistent with traveler C Z . b b Ye#concern No - See CN 3.4-45. Yes No - See CN 3.4-45. No N bS- 5 accumulator isolation isgmoved from the Applicability to the LCD- renkel-Se cl* rib/ adI.T Yo WC-1+-g7 f4-53 2'NfEAT 2. +-f3 le3 e.g le.c SA-c+ rusexr 2.+-e+ }/e No A/o A/s bc-r.ug CONVERSION COMPARISON TABLE - NUREG-1431 5/15/97

                                                     - - _ - - . - - - - - - - - - - . _ _ _ - - _ - . - _ _        ----------__-.----_-------.__.-------_-u---_---:-___---_.--------_---.--_a.. -                               . - - - . - - - - -

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

INSERT 3.4-53 Wc- 7.+-vo 7 , This change revises the ITS SR 3.4.11.2 Frequency from "18 months" to "In accordance with the Inservice Testing Program." The CTS for this surveillance establishes the frequency as being per the IST Program [ CTS SR 4.4.4.1]. - 1 i l 1 i l l

                                                       .v ...-.-          , ,

ADDITIONAL INFORMATION COVER SHEET ! ADDITIONAL INFORMATION NO: WC 3.4-008 APPLICABILITY: CA, WC REQUEST: Revise ITS SR 3.4.12.8 Bases to clarify when the SR must be performed. ATTACHED PAGES: Attachment 10, CTS 3/4.4 - ITS 3.4 Enclosure SB, page B 3.4-80 l l l l I l 1 i I

E a a vi wgawvm B 3.4.12 BASES SURVEIllMCE SR 3.4.12.8 REQUIREENTS (continued) Performance of a COT is required within 12 hours after decreasing RCS temperature to s EM RiiBk and every 31 days on each required PORV to verify and, as necessary, adjust its lift setpoint. The COT will verify the setpoint is within the PTLR allowed maximum limits in the PTLR. PORV actur.i: ion could depressurize the RCS and is not required. The 12 hour m F,;,,,.cray considers the unlike11 hood of a low temperature overpressu nt during this . w,HI ng i A Note has teb added indica ng that this isVr uired to be

  .                         set M12 hours after decreasing                      .      d leg             wc-3.6 eat temperature to s EM M Oc C" ;;n ~t k grierad until in tre L'^^       = den tre "0"A' lift ;;tpint ;;n k reduced to m a ., m m s. _ _ - u.s ,

u_ ,,,,._mm_

                           ,2, ' M,.'            LKO. .y
                                               ".y         , 'yGQ F ' ~r7 "7 'Y ~M 7

SR 3.4.12.9 Performance of a CHANNEL CALIBRATION on each required PORV l actuation channel is required every 5 months to adjust the whole l channel so that it responds and the valve opens within the ' required range and accuracy to known input. REFERENCES 1. 10 CFR 50, Appendix G.

2. Generic Letter 8811.

3. ASME, Boiler and Pressure Vessel Code, Section III.

4. FSAR, Chapter e

5. 10 CFR 50, Section 50.46.

6. 10 CFR 50, Appendix K.

7. Generic Letter 90 06.

1 (continued) MARK UP OF NUREG 1431 BASES B 3.4 80 ,, 5/15/97

ADDITIONAL INFORMATION COVER SHEET l ADDITIONAL INFORMATION NO: WC 3.4-0010 APPLICABILITY: CA, WC l REQUEST: Move the CTS list of Pressure isolation Valves to the Background Bases for ITS 3.4.14, i ATTACHED PAGES: Attachment 10, CTS 3/4.4 - ITS 3.4 Enclosure 3B, page 6 Enclosure SB, pages B 3.4-90, B 3.4-91, B 3.4-94, and B 3.4-95 l l l

m p CONVERSION COMPARISON TABLE - CURRENT TS 3/4.4 l a) (v d Page 6 of 13 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION COMANCHE PEAK WOLF CREEK CALLAWAY li)IABLO CANYON 6-01 This change adds the performance of an RCS water inventory Yes Yes Yes Yes M balance every 24 hours as a new requirement when the [ sump level detector) is inoperable. 6-02 This change allows the performance of an RCS water inventory Yes Yes Yes Yes LS-8 balance every 24 hours as an alternative to the requirement to perform 24 hour sanples of the containment atmosphere when a required radioactivity monitor is inoperable. 6-03 This change adds the word " required" to clarify that only Yes Yes Yes Yes A those detectors which are being used to satisfy the LCO must be demonstrated to be OPERABLE. No " Digital" not. Yes No " Digital" not No " Digital" not 6-04 The word " DIGITAL" has been deleted to be consistent with A the terminology used in NUREG-1431 as it relates to Channel included in CTS. included in CTS. included in CTS. Operational Tests. 6-05 This change deletes the phrase "not isolated from the No - The phrase is Yes Yes Yes A Reactor Coolant System" when referring to leakage through not part of the the SGs. current DCPP TS. 6-06 This change moves the LCO for CONTROLLED LEAKAGE (seal Yes Yes No - See No - See A injection flow) from *0perational Leakage" to LCO 3.5.5 CN 6-28-LG. CN 6-28-LG.

        "ECCS" 6-07   This change moves the listing of RCS Pressure Isolation                                                                          Yes - Moved to                                                        Yes - Moved to TRM. Yes - Moved to41 EAR-  Yes - Mov        o       -

LG Valves. FSAR. C T M- 15 4 ' 9 h it.' 1 , 6-08 Adds clarification that the valves in the RFR flow path are Yes Yes Yes Yes LS-9 not required to meet the Pressure Isolation Valve specification "when in, or during the transition to or from U /4Q Oa/er. the RtR HDDE of operation" in MODE 4. 6 09 The LCO applicability for CONTROLLED LEAKAGE (seal injection Yes Yes No - See No - See LS-10 flow) is reduced to only MODES 1.2 and 3 with the associated CN 6-28 LG. CN 6-28-LG. change in ACTION b and completion times and the associated change of the required pressure differential between the reactor coolant system pressure and the centrifugal charging pump discharge header pressure. CONVERSION COMPARISON TABLE - CURRENT TS 5/15/97

RCS PIV Leakage B 3.4.14 BASES l BACKGROUND PIVs are provided to isolate the RCS from the following typically connected systems: (continued)

a. Residual Heat Removal (M) System: i i b. Safety Injection System; and; c. Chemical and Volume Control System.

belAv: W c-2.4-0/p i The PIVs are listedVi:, th F"", Lti; . ' " . - ' 6). 2~NEEW 8 3. 4-90 i 1 Violation of this LCO could result in continuet, degradation of a l

PIV, which could lead to overpressurization of a low pressure l j system and the loss of the integrity of a fission product: barrier.

a f APPLICABLE Reference 4 identified potential intersystem LOCAs as a ! SAFETY ANALYSES significant contributor to the risk of core melt. The dominant i accident sequence in the intersystem LOCA category is the failure -

of the low pressure portion of the H System outside of j( containment. The accident is the result of a postulated failure of the PIVs, which are part of the RCPB, and the subsequent l

l

pressurization of the H System downstream of the PIVs from the ;

RCS, Because the low pressure portion of the M System is i i- typically designed for 600 psig, overpressurization failure of I the H low pressure line would result in a LOCA outside containment and subsequent risk of core melt. Reference 5 evaluated various PIV configurations, leakage testing l of the valves, and operational changes to determine the effect on i the probability of intersystem LOCAs. This study concluded that l periodic leakage testing of the PIVs can substantially reduce the

j. probability of an intersystem LOCA.

RCS PIV leakage satisfies Criterion 2 of tk OC l'clicy m a, w p . .w . x . ., LCO RCS PIV leakage is identified LEAKAGE into closed systems connected to the RCS. Isolation valve leakage is usually on the i order of drops per minute. Leakage that increases significantly 1 ( l (continued) MARK UP OF NUREG 1431 BASES B 3.4 90 5/15/97 i

e

                                                                         ._e.1 f.4 9 .et.   . ,a e PFT                     .
                                        - = E .== CCe =T      ,    ~
                                                                      = = n ~=m== = C x := :=:=

ZNSW f 2.H0 VA L'* E VALVE g) c - y, gg NUMBER SIZE (in.) FUNCTION ALI4WABLE LEAXAGE(qcm) BB8948A 10 BB89488 10 RCS Loop 1 Cold Lag SI Accu Chek 5.0 BB8948C 10 RCS Loop 2 Cold Lag SI Accu Chek 5.0 BB89480 10 RCS Loop 3 Cold Leg SI Accu Chck 5.0 BB8949A 6 RCS Loop 4 Cold Leg SI Accu Chek 5.0 BB8949B 6 RCS Loop 1 Hot Leg SI/RHR Pump Chck 3.0 BB8949C 6 RCS Loop 2 Hot Leg SI/RHR Pump Chck 3.0 BB89490 6 RCS Loop 3 Hot Leg SI/RHR Pump Chck 3.0 BBV0001 1.5 RCS Loop 4 Hot Lag SI/RHR Pump Chck 3.0 BBV0022 RCS Loop 1 Cold Leg SI/ BIT Chek 0.75 BBV0040 1.5 RCS Loop 2 Cold Leg SI/ BIT Chek 0.75

1.5 RCS Loop 3 Cold Leg SI/ BIT Chek i BBV0059 1.5 RCS Loop 4 Cold Leg SI/ BIT Chck 0.75 BBPV8702A 12 0.75 BBPVB702B 12 RCS Loop 1 Hot Lag to RRR Pumps ISO 5.0 EJ8841A 6 RCS Loop 4 Not Leg to RHR Pumps ISO 5.0 EJ8841B 6 RHR TRNS SIS Hot Lag Loop 2 Recirc 3.0 EJHV8701A 12 RHR TRNS SIS Hot Leg Loop 3 Recirc 3.0 RHR Pump A Suction ISO EJHV8701B 12 RHR Pump B Suction ISO 5.0

              \  EMV0001                         2                                                                              5.0
         \       EMV0002                         2 SI Pump A Disch to Hot Lag Loop 2 Chek                               1.0 i

EMV0003 2 SI Pump A Disch to Hot Leg Loop 3 Chck 1.0

EMV0004 2 SI Pump B Disch to Hot Lag Loop 1 Chck 1.0 4

EM8815 SI Pump B Disch to Hot Lag Loop 4 Chek 1.0 3 BIT CVCS Cut Check 1.5 EPV0010 2 EPV0020 2 SI Pumps to RCS Cold Lag Loop 1 Chck 1.0 EPV0030 SI Pumps to RCS Cold Leg Loop 2 Chek 1.0 2 EPV0040 SI Pumps to RCS Cold Leg Loop 3 Chek 1.0 2 EPS818A SI Pumps to RCS Cold Leg Loop 4 Chck 1.0 6 EP8818B RHR Pumps to RCS Cold Leg Loop 1 Chck 3.0 6 . EP8818C RHR Pumps to RCS Cold Leg Loop 2 Chek 3.0 6 EP8818D RNR Pumps to RCS Cold Leg Loop 3 Chek 3.0 6 EP8956A RHR Pumps to RCS Cold Leg Loop 4 Chck 3.0-10 SI Accu TK A out Upstream Chck EP8956B 10 5.0 SI Accu Tk B Cut Upstream Chck 5.0 EP8956C 10 SI Accu TX C Out Upstream Chck EP89560 10 5.0 SI Accu TK D out Upstream Chek 5.0 I k e

 }              CALLA"AY               U:M    1                   y/

4 m _ gang;g g

 ;        b

RCS PIV Leakage B 3.4.14 BASES LCO suggests that something is operationally wrong and corrective (continued) action must be taken. The LCO PIV leakage limit is 0.5 sps per nominal inch of valve size with a maximum limit of 5 gpe. ";; pr;;i;;; criterix ;f 1,,, fer ell ;;in ;in; 1;p;;d ;a unju;tifi;d ga;1t3 sa tM ler,.;r velv;; witM.t pic,vidia; infr,7; tia a pt;atiel v;in d;;;;d;;is ;;d .;; ult;d in .1,M. gr.,661 r;di; tim ;@;ure;.

3. ;tudy axiud;d ; ink;; ret; li;it M;;d a nin ;in n;

-g-trier t ; ;ingi; ;11:251; nin.

4 Reference-Fpermits leakage testing at a lower prc sure IA/C- I 6 o/O differential than between the specified maximian RCS pressure and the normal pressure of the connected system during RCS operation . (the maximum pressure differential) in those types of valves in which the higher service pressure will tend to diminish the overall leakage channel opening. In such cases, the observed rate may be adjusted to the maximum pressure differential by assuming leakage is directly proportional to the pressure differential to the one half power.

 '   APPLICABILITY       In MODES 1, 2, 3, and 4, this LCO applies because the PIV leakage potential is greatest when the RCS is pressurized. In MODE 4 valves in the RHR flow path are not required to meet the requirements of this LCO when in, or during the transition to or from, the RE mode of operation.

In MODES 5 and 6, leakage limits are not provided because the lower reactor coolant pressure results in a reduced potential for leakage and for a LOCA outside the containment. ACTIONS The Actions are modified by two Notes. Note 1 provides clarification that each flow path allows separate entry into a Condition. This is allowed based upon the functional independence of the flow path. Note 2 requires an evaluation of affected systems if a PIV is inoperable. The leakage may have affected system operability, or isolation of a leaking flow path with an alternate valve may have degraded the ability of the interconnected system to perform its safety furction. (continued) MARK UP OF NUREG 1431 BASES B 3.4 91 5/15/97

RCS PIV Leakage B 3.4.14 n \ (Q BASES SURVEILLANCE SR 3.4.14.1 (continued) I REQUIREMENTS f. 7 The 3/ month Frequency is consistent with 10 CFR 50.55a(g) (Ref.*&) as contained in the Inservice Testing Program, is withib#N'###

                                    ~       frequency allowed by the American Society of Mechanical Engineers (ASHE) Code, Section XI (Ref.                                              ,    and is based on the 4;c-:7.4-g/g           l

need to perform such surveillances u the conditions that l j apply during an outage and the potent 1 for an unplanned !

transient if the Surveillance were p9rformed with the reactor at i power. L4 l

                                    ~ ,. .,.. .. s a . ,... w . ~ .o. y . m . g u .. ,..
                                                                                                                                                       .      . c w w .-       ,.   .:. ;,          . .3; c. . ;              :7 ;, i, ,; . . c. o ,. v, . , a ,; x ,,;
                                     ,,<t.            ,i,J,,-          ,i,.        <.i..          .j . . t .;- ) ,- ; . .. < , -              ,c            ',
                                                                                                                                      ,.            ,3
                                                                                                                                                                   .3
                                           .~ ,         t is,.'   ....../.t.     .

,.s.,-.. '. 7

t-BASES l l SURVEILLANCE SR 3.4.14.1 (continued) REQUIREMENTS after RHR is secured and stable unit conditions and the necessary differential pressures are established. g-- Ges'Irnd ~ (SR 3.4.14.5 h ad 2 2.4.14.2 OM " V;rifying trat tre "JC ;utxie;;r; interic;;c; er; 0"C"J."LE en; urn trat "0S pr;nur; uill ret picnurin the "JC ;y;t;;;; b;yend 125t of it; d;;ign prs;ure ;f 000 p;ig. The REM EDEEl m interlock setpoint that prevents the valves from being opened is set so the actual RCS pressure must be < t psig to open the valves. This setpoint ensures the Rm design . pressure will not be exceeded and the Rm relief valves will not - a lift. x The - month Frequency is based on the need to perform the ; Surveillance under conditions that apply during a plant outage. I The E month Frequency is also acceptable based on consideration , of the design reliability (and confirming operating experience) ' of the equipment. ";;;; 0"; er; =dified by Ltc; elieuing tra "JC ;;tal;;ur; furetier, t; b; di;; bled Jar, u;ing tre "JC Sy;t;;;; ;uction reli;f v;1va fer ;;1d ;;;rpr;nur; pret;; tion in serd;re; uith 0.,0.4.12.7. , REFERENCES 1. 10 CFR 50.2.

2. 10 CFR 50.55a(c).

3. 10 CFR 50, Appendix A, Section V GDC 55.

4. WASH 1400 (NUREG 75/014), Appendix V, October 1975.

5. NUREG 0677, May 1980.

-4r- "; : a.w A;: . gg) c_g,4 g ,g g, -7v ASME', Boiler and Pressure Vessel Code, Section XI. 7,-fh- 10 CFR 50.55a(g). ._ v HARK UP OF NUREG 1431 BASES B 3.4 95 5/15/97 l-i l l JLS CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS l l CTS 6.0 - ADMINISTRATIVE CONTROLS ITS 5.0 - ADMINISTRATIVE CONTROLS RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION AND LICENSEE INITIATED ADDITIONAL CHANGES I i i f l l l I INDEX OF ADDITIONAL INFORMATION ADDITIONAL INFORMATION APPLICABILITY ENCLOSED ; NUMBER l l 5.1 -1 CA YES 5.2-1 CA, CP, DC, WC YES 5.3-1 CA, DC, WC YES 5.5-1 CA,DC YES 5.5-2 CA, CP, DC, WC YES 5.5-3 CA, CP, DC, WC YES 5.5-4 CA, CP, DC, WC YES 5.5-5 CA YES 5.5-6 CA YES 5.5-7 WC NA i 5.5-8 CA, CP, DC, WC YES l 5.5-9 CA, DC, WC YES l 5.5-10 CA, WC YES l 5.5-11 DC NA 5.5-12 WC NA 5.5-13 DC NA 5.5-14 CP NA 5.6-1 CA, CP, DC, WC YES 5.6-2 DC NA 5.7-1 CA, CP, DC, WC YES CA 5.0-002 CA YES CA 5.0-003 CA, DC, WC YES CA 5.0-004 CA YES CA 5.0-005 CA YES DC 5.0-ED DC NA DC 5.0-001 DC NA DC 5.0-002 DC NA DC 5.0-003 DC NA DC 5.0-004 DC NA TR 5.0-003 CA, CP, DC, WC YES TR 5.0-005 CP NA TR 5.0-006 CA, CP, DC, WC YES WC 5.0-ED WC NA WC 5.0-001 WC NA WC 5.0-002 WC NA WC 5.0-003 WC NA WC 5.0-004 WC NA WC 5.0-005 WC NA WC 5.0-006 WC NA JOINT LICENSING SUBCOMMITTEE METHODOLOGY FOR PROVIDING ADDITIONAL INFORMATION The following methodology is followed for submitting additional information:

1. Each licensee is submitting a separate response for each section.

2. 'If an RAI does not apply to a licensee (i.e., does not actually impact the information that defines the technical specification change for that licensee),

"NA" has been entered in the index column labeled " ENCLOSED" and no information is provided in the response for that licensee.

3. If a licensee initiated change does not apply, "NA" has been entered in the index column labeled " ENCLOSED" and no information is provided in the response for that licensee.

4. The common portions of the " Additional information Cover Sheets" are identical, except for brackets, where applicable (using the same methodology used in enclosures 3A,3B,4,6A and 6B of the conversion submittals). The list of attached pages will vary to match the licensee specific conversion submittals. A licensee's FLOG response may not address all applicable plants if there is insufficient similarity in the plant specific responses to justify their inclusion in each submittal. In those cases, the response will be prefaced with a heading such as " PLANT SPECIFIC DISCUSSION."

5. Changes are indicated using the redline / strikeout tool of Wordperfect or by using a hand markup that indicates insertions and deletions. If the area being revised is not clear, the affected portion of the page is circled. The markup techniques vary as necessary, based on the specifics of the area being changed and the complexity of the changes, to provide the clearest possible indication of the changes.

6. A marginal note (the Additional Information Number from the index) is added in the right margin of each page being changed, adjacent to the area being changed, to identify the source of each change.

l

7. Some changes are not applicable to one licensee but still require changes to the l Tables provided in Enclosures 3A,3B,4,6A, and 6B of the originallicense amendment request to reflect the changes being made by one or more of the other licensees. These changes are not included in the additional information for the licensee to which the change does not apply, as the changes are only for {

consistency, do not technically affect the request for that licensee, and are being provided in the additional information being provided by the licensees for which the change is applicable. The complete set of changes for the license amendment request will be provided in a licensing amendment request l supplement to be provided later. i J l 8.' If an NRC RAI question corresponds to a licensee initiated item, only one l information package is provided. The question number is listed in the " ENCLOSED" column of the index and both the question number and the licensee item number are listed on the " ADDITIONAL INFORMATION COVER SHEET."

9. The item numbers are formatted as follows:

[ Source][lTS Section)-[nnn) Source = Q - NRC Question CA - AmerenUE DC-PG&E WC - WCNOC CP - TU Electric TR - Traveler ITS Section = The ITS section associated with the item (e.g.,3.3). If all sections are potentially impacted by a broad change or set of changes, "ALL" is used for the section number. nnn = a three digit sequential number l i ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 5.1-1 APPLICABILITY: CA REQUEST: CTS 6.1.1, ITS 5.1.1, Change 01-01-A and Difference 5.1-2 (Callaway) Comment: Difference 5.1-2 states that the STS is revised to " maintain CTS." However, given that Change 01-01-A Insert 1 includes new language into the CTS it is unclear how the CTS is being maintained. This addition of new language into the CTS and I deviation from the STS is not justified. Provide justification. l FLOG RESPONSE: l The new language that was added to the CTS in Section 6.1.1 was added based on comments from our Onsite Review Committee (ORC). This added language currently resides in CTS Section 6.5.3.1c but this information will be moved to Section 5.6.2 of the OQAM upon implementation of the ITS. The ORC decided that it was appropriate to add this information into Section 6.1.1 of the CTS as well as the OQAM. Likewise the deviation from the STS in ITS 5.1.1 is justified since we are maintaining CTS by adding I the same information that is currently in the CTS Section 6.5.3.1c to the CTS Section 6.1.1 as well as to ITS 5.1.1. ATTACHED PAGES: None i l 1 ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 5.2-1 APPLICABILITY: DC, CP, WC, CA-REQUEST: STS 5.2.2 b and Difference 5.2-2 Comment: TSTF-121 has been withdrawn for modification, combination and resubmission. Use current ITS. FLOG RESPONSE: Traveler TSTF-258 has been submitted to the NRC for review. This traveler superseded travelers, TSTF-86, TSTF-121, and TSTF-167. TSTF-258 is based on the recommendations in the April 9,1997 letter from C. Grimes (NRC) to J. Davis (NEI), with some exceptions. The FLOG submittals have been revised to incorporate TSTF-258. The latest industry status on TSTF-258 is that the NRC has requested changes to Section 5.7, High Radiation Area. See response to Comment Number 5.7-1 for how the FLOG has addressed the NRC comments on TSTF-258. ATTACHED PAGES: Encl. 2 6-2, 6-5, 6-6, 6-18, 6-21, 6-24 Encl. 3A 2,3,5,6,7,9,10 Encl. 3B 2,4,5,7 Encl. 4 1, New LS-5 Encl. 5A Traveler Status page, 5.0-3,5.0-4,5.0-5,5.0-9,5.0-10,5.0-33,5.0-39, I 5.0-40 Encl. 6A 1,2,4 Encl. 6B 1,2,3,4 l l 1 1 l 1 ADMINISTRATIVE CONTROLS _- , An e gv 7enent o pe yahor skutt be ass q'nea when 6.2 ORGANIZATION (Continued) .p y ;,,,n g yt cgr c,36 yn ,,g4,mgep, ( UNIT STAFF oPtystof S ha ll Ir.e usiO n44 wh e n Hw. U M is # n / N\ ODE 1,1, 3 , o r 4 . / 6.2.2 Th Unit organization shall be subject to the following: Ol LG, ,+ , _,c.+.

g. E -. _ L. . m . m. . A m. +. .

m, e.

b. 4 f +. e .h. . ., ,. 1 k. .m m...r. m m..m m . e n ..s.. m .f. ... . . ,.. . _L. e. f. .

+ k. m....4._.:.._.~.._ . i 4. . . . . m . m .m , 4. .+. 4. m. m .-r . ... , w m. m.m T. _e .. s , m. c_.9_3_ ., 01-06-LG,

b. At it::t on^ licen: d Op^rator : hall b^ i,n th ^ntrcl room wh:n 4,

r..._,. 2. ... _ . .. _m.,,.m.

t. _. . . .. . . , , , a. s. 4. +. 4. m. m. . , ms4.e. + s.e_

,,,4.+_ 4 _, u.n.n e_1, .m, . ., ,, ,___. ___ , , _ ._. _ _._.; .e ._ .m.. :. .m_. nmm_,+ ... .... .r . ..m. .. .... . . .. ,w,,, um 4, ,m, .. x. . .. o[.05- A centr:1 reem; 4 . m technst uen

c. An indi idu 1 frcr the fealth jhysics .rg ri:: tier", qu:lified i-- O h 02- A

-sed 4ctier prct:: tier precedure;, shall be on site when fuel is in the reactor; ,,, rmme .,w- ,y _t...,.,. um ,__ m..m .a . m. .a.a4. --+,m , mm..m4..,ms su.

d. n,, . c. m . n,m

. s.nne . . ~ _ .m u. . . , 4,mmees ec,4nr n,e ....., m_r ,+m. -.s . . . u. . i s . . 3 4. m. m. m.. . m s., e . . 4. m. .. nmm..,+m. -r - - - - m. -r----- Limited to Tuci Handli gn "h 5:: n -- ether conc"rrant responr4b41- O l- 03-A ...m.. sm. :. . ,m +..L. 4. ., e .rm m. . .., + 4. .m m. . ., ee 2_m . ... . a_ r e. _i . n _. :s _ yu_ m; ._ _ v r. .. . .., .-....c:.._ i.v. m __. . . . u. ..om _ , , .. ,L..,,., u.. m . .., 4.. m. +. ., 4 m. mJs .m.., . 4. +. m .. ++ 1. , + 4 . m~a r .. T h. e. E4.m . e r.a. n ., s m .r h. ., 3 1 m. m_ +_ 4 m. r. i. n_ s_+. m_ L. . m... eL4t+. g g,og, gg ~4 imur :F 4 #t crc'.'

"u;;r i:cr and th^ tuc cther :mber: ef the n^ ^:::ry fu : f: :hutdown Of the unit :nd any por nncl r: quired f r Other ::: rti:1 function: during "r :: rgency;

f. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related i functions; e.g., licensed Senior Operators, licensed Operators, !

h2alth physics _ personnel, eguipment o'perators, and key maintenance personnel, 'pse In r t 2- 2 a. )) s 6? 5'2 - 1 If -The amount of ev^rtims worked by Unit st:ff m^mber: performing -stfety-re4tted function: th:l' be 'i-ited in accordance uith the l l _ oq_ A MC-Pol-ity-Steement-on--working hour: (Centric Letter NO. 82-- 12 ) ; i -tM- . 9 The Superintendent, Operations, Shift Supervi cr:, and Operating Sue crvi:cr; shall hold a senior reactor operator license. The Unit Ai.L v1 i,A A G l nuc - _i.__ n____.__ .r.. .o,, _L . . .L..m.,.A. ,. m . ., r. ... . m . ,. m m .r . ., + . .m. . , 4. ,. m. m. . ., m. m.,. .,. m., m.4. m_ r. _ ...... mm. ..m.. vy.. ... .,:. . m a , m. s..4. e N polika$n IU W bt NACA #May b; l^:: 15:n th: 4,imur requirem:nt:kfor a period of time not to exceed 2 hours in order to accommodate unexpected absence provided im::.ediate action is . . . ol-ol- A taken to fill the required positionK. CALLAWAY - UNii 1 6-2 Amendment No. 77, 86 Correction Letter of 9/21/94 INSERT 2-2a Q 5.2-1 l The controls shallinclude guidelines on working hours that ensure adequate shift coverage shall l be rnalntained without routine heavy use of overtime. Any deviation from the above guidelines i shall be authorized in advance by the { Plant Manager } or the { Plant Manager's ~ J @ nee, in accordance with approved administrative procedures, and with documentation oi s an9 Tor granting the deviation. Routine deviation from the working hour guidelines shall t' - authorized. f l l I l l l l l' l j l l ( .-. -_ . ~ . - _ . - ~. .. . - - - 1 1 1 * ,..e .t f. . . f . *) i. l l m , .,m..m - . . - ,SN. _ ,.mm -O l- % -LG. l n a n a nvr. .a n a r . L vuorm va m . 4. . 4, mvn. . - ! 1 l mn,e .,, ,.nmu m . nUMBE-OF-IND.Y h....o . . . . . . . . . . . . . 1 r nc i --nw i v r a u. rua 4 : 4 vn l unne , ru wu ., , e via _- ,. unne , __ <m us , v. .r oo ., .,. - eaa , -r~ m___ -. ,i. n o . .. nu & -m - av & . EIO s. .m.._ _ _. . b 55 Sh4ft-Supe,rvisor-with-e-Seftier-Operator licen:t On Unit 1 . nv.--- ,_

46. N ue . . . , . u.

,b. ,e _ _ 2. _W s_ n.y s__..__ . p k. , , ,,_____ __ n_ , en m - ,ed4Heua4-e. t x.. U.. . n. r . . . . . . . b s . 6, s ,4.-......... ...n. ...,..2... . en .. .4... kV -- e... .- n.__..__ bMW VVE. .bV. STA - Sh+f4-TechMc:1 Adviser-- t o crR so Wmlb.}G) W 4.2 2.n. The Shift Crew Composition may be one less than the minimum requirements of I -Teble 0.b; -for a period of time not to exceed 2 hours in order to accommodate OI-%-LG 3 unexpected absence of on-duty shift crew members provided immediate action __ taken to restore the Shift Crew Composition to within the minimum requir _ s, t. u., s, ,_e.,.,.. s, '

r. 4s 4 _ 2._._.. ... _ _ . r. , __ _. ,g ._.. i

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.s .s s.. . . u. .e 4 I r s.s k i -. n.m...., 34 . .. Ol- 02.- A sh li :: d :ignated-to--anume-the-centrol ree 00- .:nd fun: tier. I \ .n... _ con,. ..4.. .. w . .. .. e. .u. ,. . .e... .. ., .. .a. . _ _ n . . . .......r.... 4... . , .e.r... . . . a. ...._ _. 0(-03-A _.m..._ .e .m rw. u . o _,m,, m oo.. u. ...onn w .n osv&., _ >_ mmm-, , ., , __> ,, u, s, onw ,. un. ,, uns ._,___ 7m_- e ,

u. .s n

3 -Superviser Or--the---individu:1 vith-a-Senior Operator ' ice se .::ts the ol-06-LG ( que44.gcat.4., , . ... < . ._ .6.. . . .e, n. . . . . .,. . .a.. . a. w. u, + w. m o.r . .. .... j ~ } Q 5.2-(" CALLAWAY - UNIT 1 6-5 Amendment No.71 . l 1 1 jDMINISTRATfVECONTROLS l 1 l -5.2.3 INDE"ENDENT-SAFETY ENCINEERINC CROUP f!SE44-O l-04-LC, I r,,iu - m,- J -6.2.3.1 The ISfHHhall function to examine-plant Oper:t4ng-char +cteri: tit +, l ""C i:suance:, i de n ,t y :dvi+er4es, REPORTABLE EN"TS :nd Other :: gree:

( phnt-des 4gn--and Operat4ng exper4ence-infc. ,: tion, inc4eding phnt of imihr d :f gn, whiebmay-indicate-area: for i; proving phat ::faty. The 1500

-sham-make-deta44ed-recenmendat40ns-for-revi:ed precedere:, :quipm:nt ;;difica- -tion:, =:4etenance-ac4444t4 :, Operat4:n: ::tivitie: er Other ::::: of impr-oving plant safety te the Manager, Oe:11ty A::er:nt: :nd th: M:::g:r, l -Ge44eway Phnt. . ~ ~ . . , - w""""'"" Ol-04-LC, 0.2.3.2 The SEc : hail bc :;;p::cd of-et h::t five, dedi::ted, fe!'ti- -en94*eers-located en :ite. E;;h :h:11 h:v: ; b:cheh r': d:gre: in engin:: ring er rehted ::iene: :nd :t 10::t 2 y r: profess 4:::1 hvel enperience in his field. RESPONSIBILITIES ol-04-LG 5.2.3.3 The ISEC :h:11 be re: pens 4b1: for ::intai"ng :ervei44:nce of phnt i

tivi 44es-to provide-independent verification

that the:; ectivitie: ;re i performed-correctly end that hum;n errer: ere reduced e: auch E: precticel. i

~~ Ol-0 4 L G , 5.2.3.' Da erds of activities perfe =ad by tha -!SEC shall be prepared, -ma4*te4aed, ;nd forwarded each-ealender-month-to-the M:n:g:r, Ou:lity A::gr:::: l end the ".en;ger, C;1hw:y Phnt. 7 ! 6.2.4 SHIFT TECHNICAL ADVISOR An80,Wd d l 1 The Shift Technic;l Advi: r (STA)** shall provide technical support to the 01-l5- A ! ) Shift Supervi:cr 'n the areas of thermal hydraulics, reector engineering and / plant analysis wi h regard to the safe coeration of the _ unit. Q5.'l-Ik ) bU # ( 6.3 UNIT STAFF 00ALIFICATIONS __ 6.3.1 Each member of the unit staff shall meet or exceed the minimum > qualifications of ANSI /ANS 3.1-1978 with the following exceptions: 6.3.1.1 Shift Supervisors, Operating Supervisors, Reactor Operators, and Shift Technical Advisors shall meet or exceed the qualifications of ANSI /ANS 3.1-1981 as endorsed by Reg. Guide 1.8, Revision 2, with the same exceptions as contained in the current revision to the Operator Licensing Examiner Standards, NUREG-1021, ES-202. 1 1 6.3.1.2 The Radiation Protection Manager shall be a supervisor with line responsibility for operational health physics who meets or exceeds the qualifications of USNRC Regulatory Guide 1.8, September 1975, for a Radiation Protection Manager. The Radiation Protection Manager will be designated by the Plant M . M ' Notre:fonsidefor:igneftfunction. Ol-04-LG **The STA po 'dion shall be manned in MODES 1, 2, 3 and 4 unless the Shift Supe or n with a Senior Operator license meets the OH5- A qualificatio s for th; STA a required by the NRC. u _; CALLAWAY - UNIT 1 6-6 Amendment No. 6b GG,107 .- =-. _ . . . J ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) \ I e. Radioactive Effluent Controls Proaram (Continued) ] ( funchknAl c@.mtO

6) Limitations on thet:per:::mty ana use of the liquid and 07.-os-A gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming t Q ppendix I to 10 CFR Part 50, W si@ de ;

A 5.2-\\ \

7) Limitations on the dose ate resultingl from radioactive I '

r and the SITE s effluentsYto areas be dh material released in gaseBOUNDARY terfer 'ag to the ..r 4 -50, Appendix 0, T;ble II, Colurc.n 1, de:::Y:Mei ted ud 02-ig A

8) Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,

9) Limitations on the annual and' quarterly doses to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives creater tha 8 days in ga 61Ts effluents released to areas beyond the SITElc?5.2-d BOUNDARY co forming._to Appendix I to 10 CFR Part 50,

, b.yo4 w site (souwoawD 02A9-A o the annual dose or dose commitment to any MEMBER i 10) Limitatio OF THE PU IC due to releases of radioactivity and to radiation / from urani fuel cycle sources conforming to 40 CFR Part 190. j

f. Radic1ccical Ervircreert:1 uc Hterine Prec 02-oG-A A program : hall bc provided to acnitw-the r: diction :M r:dienaclide: in the enviren: cf the-phnt . The progrs= :h:11 provide (1) repec:entative me;;urc=nt: cf r;dicactivity in the-hight:t pctential expe:ure p thw:y:, =d (2) verification of the- .

---accuracy of the ef#'eent meritering progr= =d =cdeli .g Of i environment +1-exposure p;thway:. The progr= sh:11 (1) be cont:ined ! in the CDCM, (2) confer n to the guidance of Appendix ! to 10 CIR I Fert 50, end (3) include the feliswing. l

1) ManitcKng, :=pl4ng, =: lyric, =d reperting of : :diation 2nd -

radienuclides in the environment-in-acccrdance with the 'methodelegy and per;;cter: in the 00CM,

2) A Land Use Ca:u: to en:ure that ch = ge: in the ucc of-areas-at

-and beyond the SITE SOUNDARY arc identified and that- -+edW4et4en te the enitor4ng progr= are nde if required by the re: ult: cf thi: cense:, :nd 02-22-A ,., it) 1h provuions of Tec%cd sgcJic h n 4.o.1 c4 4 0 3 l ) Mt appluaW h W Rdkleyc(l Ehn4 Conblr Proyam. QS.2-l , CALLAWAY - UNIT 1 6-18 Amendment No. -5g10 3 r\ s ey t 1 A sh al Ee. in accor6 nct Uh -M, I .. D - -a) m Q 5.2-1 , tha SIT: 30'J"0A:1Y :h:11 b: li=it

1. For noble gases: Less than or equal to a dose rate of 500 mrems/yr to the total body and less than or equal to a dose rate of 3000 mrems/yr to the skin, and

,g

2. For iodine-131, iodine-133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to a dose rate of 1500 mrems/yr to any organ; l

i l l l l ADv.IN15iRATIVE CONTR0L5 THLY OPERATING REPORT _ 7- , 6.9.1.8 Routine reports of operating statistics and shutdown experience, including de: .ertatier of al' challence; to the pressurim- PORVi u, RC3 03-l W - :: f e t; v3'ves.J5 hall be submitted on a monthly basis to thi Directgr, Offict cf Res;;r;c Management, U. S. Nucicar Regulatory Coristicr., L'::Hngton, D. C.g3.og g 20:55, with : ;;py to th OC Rcgional Of'i;;, no later than the 15th of each month following the calendar month covered by the report. CORE OPERATING LIMITS REPORT 6.9.1.9 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT prior to each reload cycle, or prior to any remaining portion of a reload cycle, for the following:

a. Moderator Temperature Coefficient SOL :nd ECL limits and 20 surveillance limit for Specification 3/4.1.1.3, p;- g_g g

b. Shutdown Bank Insertion Limit for Specification 3/4.1.3.5, ,

c. Control Bank Insertion Limits for Specification 3/4.1.3.6,

d. Axial Flux Difference Limits, RAFD0 target band, and APL'40 for l .

Specification 3/4.2.1, l i I RTP l e. Heat Flux Hot Channel Factor, Fg , K(Z), W(Z)ND, APL"O and W(2) MFD 0 (as required) for Specification 3/4.2.c,

f. Nuclear Enthalpy Rise Hot Channel Factor F,g. and Power l Factor Multiplier, PF H, limits for Specification 3/4.2.3. 03-l 4-M y

Tne analytical methods used to determine the core operating limits shall be 63-15-M those previously reviewed and approved by the NRC, specifically those described , j in the following documents. l a. WCAP-9272-P- A, " WESTINGHOUSE RELOAD SAFETY EVALVATION METHODOLOGY", l July 1985 (W Proprietary). (Methodelegy #cr Speci#icatica 3.1 1.3 Moderate- Temper:ture-Ccefficient; 3.'.3.E Sr.utcom S nk Inscrtion Limit; 3.1.J.; N 'O-A , Control Cenk Insertion Limit, :nd 2.2.0 L:lc;r :nthe4py-i Rise "st Chennel Fecto .) l

3. %Ab.m Mupo Limas Sor SpEJddian 3/41.1, o.nb

h. Re.bdin) Mn C nc.edreon GenN b he'cdMdn 3/4 0 i ,

t CALLAWAY - UNIT I 6-21 Amendment No. 28,50,FS,72 - ] ADMINIS~RA*IVE CONTROLS l 6.12 HIGH RADIATION AREA (Continued)

b. A radiation monitoring device which continuously integrates the l l radiation dose rate in the area and alarms when a pre-set integrated 4

dose is received. Entry into such areas with this monitoring device l may be made after the dose rate levels in the area have been estab- l 11shed and personnel have been made knowledgeable of then, or ~

c. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible fo,-

providing positive control over the activities within the area and shall perfem naciodic radiation surveillance at the frecuency specified E " lth "hy:ie: : :: m t ge m, anal]In the RWP." or conenvously pried (grp(6/fr/r toi <tud 45) (s ocm) i 6.12. 'Tn addition a me r irements- speci tion 6.12 areas accessible to personnel with radiat levels great than 000mR/ hat (10 [4::in. ) from 03-H-A the radiation source or rom c.ny surfac ieh the radiation penetrates shall be provided with locked doors to preven <= :2 2 :ri::dTentry, and the keys shall g. be maintained under the administrative control of the Shift Supervisor / i Operating Supervisor on duty and/or health physics supervision. Doors shall Q 5.1-0 remain locked except during periods of access by personnel under an approved RWP which shall specify the dose rate levels in the imediate work areas and ge, che maximum allowable stay time for individuals in that area. In lieu of the gg _% 4 stay time specification of the RWP, direct or remote (such as closed-circuit j TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area. 4 ch 30 69 ! For individual high radia on areas accessible to personnel with radiation levels

of greater than 1000 mR/h that are located within large areas, such as PWR 03.[1- A containment, where no enclosure exists for purposes of locking, and where no enclosure can be reasonably constructed around the individual area, that indi-

. vidual area shall be barricaded, conspicuously posted, and a flashing light shall be activated as a warning device. C.13 PROCESS CON'*ROL PROGRA" '?CPF 0 3-l2-LC. --Ghanges to the POP; -4. Ball be d0oumented :nd rec rd: Of review: pe # : :d 0h:11 bc -eeta4ned-as-requ4 red by Specification 5.10.20. '"hi: deemca te.- - --- tion : hall cata4et-

1) Suf44e4ent-4ef : tien :: support-the-ch:ngs together with

-the-4ppropr4ete-analyses Or ev:htthn: jus 4tfy4ng the -change (s), and

2) A determinat4on-that-the-change-wi44-maint+4*-the Over:11

---conformance-of the :clidi#hd weste product t exi: ting- -requ4rements-of-feder44r-State, Or- Other Opplict.ble regula- -t4ensr CALLAWAY - UNIT 1 6-24 teendment No. 50 - ., ,,_,- o -- - - y, ... m,.m . m _.#. w_..~ m - - - - CHANGE NUMBER NSEC DESCRIPT10N Criterion II and V: ANSI N18.7 1976: and N45.2 1971). Therefore, du".ication of these requirements in the CTS is not required. 01 05 A The requirement for the presence of a Reactor Operator (RO) or an SR0 in the control room is deleted from the TS since the requirement is consistent with and duplicative of the manning requirement in 10 CFR 50.54(m)(2)(iii). I Deletion of the CTS requirements does not change the manning requirements and is therefore considered an administrative change. 01 06 LG The details regarding the minimum shift crew requirements have been removed from the CTS because they are redundant ) to 10 CFR 50.54(k), (1), and (m) with the exception of the l requirement for non-licensed operators. The corresponding j ITS section 5.2.2b requires meeting the requirements of these regulations which specify the shift complement regarding licensed operators for all modes of operation. The minimum shift crew requirements will be moved to a licensee controlled document. 01 07 LG Revises Section 6.2.2a, Unit Staff Organization, to reflect the non licensed operator staffing requirements for a single unit site consistent with NUREG 1431, Rev.1 requirements. The minimum shift crew composition as described in Table 6.21 has been moved to a licensee control document and provides requirements for the minimum number of non-licensed operators necessary for plant operations. This proposed change is consistent with NUREG 1431. Rev. 1. 01 08 LG Hove the fire brigade requirements to a licensee l controlled document. Moving these requirements is consistent with NUREG 1431, Rev. 1. These requirements can be found in BTP ASB 9.5-1 and their duplication in the ITS is not required. 01-09 Af A Net und. 1Me f 6 3 A -2A Qy.2.-1] 01-10 M Not applicable to Callaway Plant. See Conversion Comparison Table (Enclosure 3B). 01 11 A Not applicable to Callaway Plant. See Conversion Comparison Table (Enclosure 3B). 01 12 A Not applicable to Callaway Plant. See Conversion Comparison Table (Enclosure 3B). DESCRIPTION OF CHANGES TO CURRENT TS 2 5/15/97 INSERT 3A 2a Q 5.2 1 i 1-09-A The CTS requirements concerning overtime being in accordance with the NRC Policy Statement is replaced by referring to administrative procedures for the control of working hours. The proposed change provides reasonable assurance that safe plant operation will not be Jeopardized by impaired performance caused by excessive working hours. Specific controls for working hours of reactor plant staff are described in procedures that require a deliberate decision making process to minimize the potential for impaired personnel performance, and that established l procedure control processes will provide sufficient controls for changes to that I procedure. Replacement of the CTS reference to referring to administrative controls does not change the requirements associated with working hours and is therefore considered an administrative change. These changes are consistent with the NUREG-1431 as modified by TSTF-258. l l 4 i CHANGE NUMBER H$liC DESCRIPTION 01 13 A Not applicable to Callaway Plant. See Conversion Comparison Table (Enclosure 3B). 01-14 A Revise this Section, to delete the shift supervisors and operating supervisors from this section as required to hold a senior reactor operator license. Shift supervisors and operating supervisors fulfilling the staffing requirements of 10 CFR 50.54(m)(2)(i) are required to be licensed under 10 CFR 55. This change is consistent with ! % NUREG-1431. Revision 1. (OI- 15 A Jnse et 3g.3Q { Q 5.2-M 02-01 A CTS Section [6.6.la] for Reportable Event actions has been deleted from the CTS. This section only repeats the regulatory reporting requirements defined in 10 CFR 50.72 and 10 CFR 50.73, and is unnecessary in the TS. Deletion I of this section from CTS does not impact safety because it l is redundant to the regulations cited and is therefore l acceptable. l 02 02 LS 4 CTS Section [6.7], Safety Limit Violation " requirements to notify the NRC within 1 hour following a violation of a safety limit (SL), submit a Safety Limit Violation Report and not resume plant operation until authorized by the Commission are being deleted. These requirements are a 1 duplication of 10 CFR 50.36(c)(1), 10 CFR 50.72 and 10 CFR 50.73. [The 14 day safety limit violation report is not required since 10CFR50.73 would require a 30 day LER.] Since the plant must meet the applicable requirements contained in the regulations, sufficient regulatory controls are maintained to allow removing these i duplicate regulatory requirements from the current TS. The notification requirement to executive management and l the review comittees is an after the fact notification and is not necessary to assure safe operation of the , facility. As, such this requiremen.t is not necessary to l be included in the TS. These changes are consistent with NUREG-1431 and traveler TSTF-5. 02 03 A The implementation procedure requirements related to [the Security plan.] the process control programs and the radiological environmental and offsite dose calculation programs are deleted from the CTS consistent with NUREG 1431. These types of procedures are either required by Regulatory Guide 1.33, Rev. 2, Feb. 1978 referenced in the ITS or the CTS. covered under the provisions of ITS 5.4.1.e or required by 10 CFR 50, Appendix E and 10 CFR 50.54(p) and (q). Therefore, these requirements are duplicative and unnecessary. DESCRIPTION OF CHANGES TO CURRENT TS 3 5/15/97 INSERT 3A-3a Q 5.2-1 1-15-A This change revises the CTS to eliminate the title of" Shift Technical Advisor (STA)." STAS are not used at all plants (the function may be fulfilled by one of the other on-shift individuals). This Section is revised so that it does not imply that the STA and the Shift Supervisor must be different individuals. Option 1 of the Commission Policy Statement on Engineering Expertise on Shift is satisfied by assigning an individual with specified educational qualifications to each operating crew as one of the SROs (preferably the Shift Supervisor) required by 10 CFR 50.54(m)(2)(l) to provide the technical expertise on shift. Eliminating the title of STA is considered an administrative change since the requirement for engineering expertise on shift is maintained. This change is consistent with NUREG-1431 as modified by TSTF-258. l l 1 l l L l l CHANGE WEIR HSHC DESCRIPTION 02 10 M Not applicable to Callaway Plant. See Conversion l Comparison Table (Enclosure 3B). ) 02 11 M New program requircments, " Safety Function Determination Program" and " Bases Control Program" would be added. , consistent with NUREG 1431. Although these new programs l I reflect current plant practice. delineating them in the ) ITS would be more restrictive. ' 02 12 LG Move the Emergency Diesel Generator Reliability Program requirement to a licensee controlled document. Moving this program is consistent with NUREG 1431. Rev. 1. 1 02 13 LG Revises Section 6.14 item b to move the requirement that 1 ODCM (or similar programs and procedures) changes require l review and acceptance by onsite review committees to the 00CM. The onsite review of ODCM changes is currently required per [ procedures]. This change is consistent with NUREG 1431. i 02 14 M Per GL 89 01, concentrations of radioactive material releases in liquid effluents to unrestricted area shall conform to 10 times the concentration values in Appendix B, Table 2. Column 2 of 10 CFR 20.100120.2401. { Prepcxd trivekr pcrdirt. %64crt 3A-5D I 05.2.-[ \ 02 15 LG CTS Section [6.6.1b] contains requirements for the plant review and submittal of a reportable event. This ; information is to be moved to a licensee controlled document. Given that these reviews and submittal of results are required following the event without a specified completion time, the requirements are not necessary. The moving of this information maintains consistency with NUREG 1431. Rev. 1. 02 16 A To maintain consistency with the Bases for TS 3/4.8.1, change the Diesel Fuel Oil-Testing Program description for sampled properties of new fuel oil from "within limits" to " analyzed" within 30 days following sampling and addition of the fuel oil to storage tanks. This wording more clearly defines that within 30 days following the initial new fuel oil sample, the fuel oil is analyzed to establish that the other properties specified in Table 1 of ASTM D975 81 are met. This change is consistent with the Bases ! for SR 3.8.3.3 of NUREG-1431. The " Reactor Coolant Pump F TsTF O'l] heel Inspe dion Program" is l 2 17 LS 1 revised consistent with WOC 4 The proposed changes to provide an exception to the examination requirements in Regulatory Guide 1.14. Rev.1. " Reactor Coolant Pump (Q 5 5-l} DESCRIPTION OF CHANGES TO CURRENT TS 5 5/15/97 INSERT 3A-Sa Q 5.2-1 ! This limitation provides reasonable assurance that the levels of radioactive materials in bodies of water in unrestricted areas will result in exposures within (1) the Section ll.A design objectives of l Appendix I to 10 CFR Part 50 and (2) restrictions authorized by 10 CFR 20.1301(e). These changes are consistent with NUREG-1431 as modified by TSTF-258. l l I I i l CHANGE NUMBER HS!1C DESCRfPTION Flywheel Integrity." The proposed exception to the recommendations of Regulatory Position C.4.b would allow for an acceptable inspection method of either an ultrasonic volumetric, or surface examination. The acceptable inspection method would be conducted at ten year intervals coinciding with the Inservice Inspection schedule required by ASME Section XI. This change is consistent with the NRC Safety Evaluation Report associated with WCAP 14535 Topical Report on Reactor Coolant Pump Flywheel Inspection Elimination. Insert 3A-M 02 18 A Revises the Radioactive Effluent Controls ogram dose rate limits released to areas beyond the _ ite boundary to pT.2.-l} reflect new 10 CFR Part 20 requirements.Ttonsisten with FNRC lettAr datep 7/ZB/95/(Christopher I. Grimes Owners t Groups). Addifionally,/the NRC /ssued a draft rTt-- - Lette 'r,199 whic.h p upused C anwes to the-S dard Tech ical S cificati ns. Thi change is con ' stent With l the draft neric Le ter and REG-1431. Rev. 1 as amende by a propcped b ave r er to ref ect p cherses co ps4 stent wi I e1 CFR Part 20. L _ 02 19 LS 2 Consistent with NUREG 1431, the surveillance interval for verifying that other properties are with limits for ASTM 2D fuel oil is changed from "within 30 days" to I "within 31 days" after obtaining a sample. The fuel properties that can have an imediate detrimental impact on diesel combustion, (i.e., API gravity, kinematic viscosity, flash point and appearance) are verified prior to addition to the storage tank. The "other properties" may be analyzed after addition to the tank. The 31 day verification interval for these properties is acceptable because the fuel properties of interest, even if they are not within their stated limits, would not have an imediate affect on diesel generator operation. The CTS 4 30 day verification interval was probably chosen because it was a convenient time interval for sending the sample i and receiving the results from the laboratory selected for i testing. NUREG 1431 has selected a 31 day testing ) interval. The 1 day increase in the interval would not j have a significant affect on the acceptability of the i diesel fuel oil. i 02 20 A Consistent with NUREG 1431. Rev. I and traveler TSTF-118, i add the statement that the provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing ' Program frequencies. These sentences provide consistency with the current application of these requirements as provided in ISTS 5.5.6 and ISTS 5.5.11. Amendment [112] moved the Diesel Fuel Oil Testing Program to Section

6.0 DESCRIPTION

OF CHANGES TO CURRENT TS 6 5/15/97

_.. - - - ._. - - .-~ - . . ~ - - _ _ _ - . - - . . - . . - . . - . - INSERT 3A-6a Q 5.21 After issuance of Generic Letter 89 01,10 CFR 20 was updated. The NRC issued a draft Generic Letter,93-XX, on proposed char,ges to STS NUREGS based on the new 10 CFR 20. The proposed changes are consistent with the draft generic letter, the April 9,1997 letter from C. Grimes to J. Davis (with some exceptions). The proposed changes maintain the same overall level of effluent control while retaining the operational flexibility that exists with current TS under the previous 10 ) CFR 20. These changes are intended to eliminate possible confusion or improper implementation of the revised 10 CFR 20 requirements. The proposed changes are consistent with NUREG-1431 1 2 as modified by TSTF 258. ;

't

't 4 1 l l

I CHANGE l NUMBER ! gig DESCRIPTION of the current TS but did not include a statement that the provisions of SR 3.0.2 and SR 3.0.3 are applicable. SR 3.0.2 and SR 3.0.3 are applicable to the surveillances which reference these programs, and therefore, the lack of gg g .3bg an applicability statement in the Programs introduces _ l gy q confusion. < 03 01 A Revises " Routine Reports" section to be consistent with NUREG 1431. The method for submitting all reports is revised to be in accordance with 10 CFR 50.4. Since this change merely makes the TS consistent with the regulations, it is considered administrative. 03-02 A The requirement to submit a Startup Report is deleted from the CTS to be consistent with NUREG 1431. This report required no staff approval and was submitted after the fact and is therefore not required to ensure safe plant operation. The approved 10 CFR 50, Appendix B, QA Plan, and FSAR startup testing program provides assurance that ; the affected activities are adequately performed and that l appropriate corrective actions, if required, are taken. 03 03 A The Annual Reports section is revised to be consistent with NUREG 1431 and traveler TSTF 152. Names and formats are revised consistent with NUREG 1431. Also, revises the l Annual Report section to reflect the new 10 CFR Part 20 requirements and associated recommended changes noted in NRC letter dated July 28, 1995 " Changes to Technical Specifications Resulting from 10 CFR 20 and 50.36a Changes." (From Christopher I. Grimes to Owners Groups Chairs) 03 04 A The requirement to report specific activity limit violations is deleted consistent with NUREG 1431 and traveler TSTF 152. This report is a history of RCS specific activity LC0 entries. GL 83 43 and revised reporting requirements in the regulations intended that LC0 entry reports no longer be required. The reporting requirements in regulations cover situations such as seriously degraded barriers (fuel failure). Therefore, every violation of the RCS specific activity LC0 need not be reported. Serious degradation of a fission product barrier, among other more serious events are required to be reported by 10 CFR 50.73. This change is administrative in that it only affects reports and do not affect plant operations. 03 05 A Not applicable to Callaway Plant. See Conversion Comparison Table (Enclosure 3B). DESCRIPTION OF CHANGES TO CURRENT TS 7 5/15/97

_ __ _ __. _ _. .- _ _ __ _ _. _ _ _ . . .- _ _ . . _ . . _ __ ._ ~ ___ - .___. _ . _ . 1 INSERT 3A-7a Q 5.2-1 l 02-22 A The Radioactive Effluents Controla Program is revised to include clarification statements denoting that the provisions of CTS 4.0.2 and 4.0.3 are applicable to these activities. These statements of applicability clarify the allowance for

surveillance frequency extensions and allowance to perform missed surviellances. Generic Letter 89-01," Implementation of Programmatic Controls for Radiological Effluent Technical Specifications and the Relocation of Details of RETS to the Offsite Dose Calculation Manual or the Process Control Program" allowed licensees to relocate the Radiological Effluent Technical Specifications and establish the Radioactive Effluents Control Program in the Administrative Controls Section of the Technical Specifications. This change effectively implements the CTS requirements that were relocated per Generic Letter 98-01.

This change is considered an administrative change since the changes are in the presentation method only. This change is consistent with NUREG-1431 as modified by TSTF-258. t l l l I i l 4 i i

CHANGE NUMBER 16E DESCRIPTION

   )                                 required by 10 CFR 20.1101(c). The CTS is redundant to requirements in the regulations and thus is deleted.

The high ra etion ar is rev 'd to consi nt w 03 11 A NUREG 1 and th ew Part requ' ements Chan s are non- chnical add cla ficati and co orm h IWR G 1431 d RG 8.3 . .f Me< 6 3 A 9 b p c,,i_ l ( l I 03 12 LG The Process Control Program (PCP) section is proposed to l be moved outside the CTS consistent with NUREG 1431. The PCP implements the requirements of 10 CFR 20, 10 CFR 61, ! and 10 CFR 71. Therefore, relocation of the descript.'on

                                     'of the PCP from the CTS does not affect the safe operation                             ;

of the facility. The PCP may be adequately described in licensee controlled documents. 03-13 H The following report [s] will be added to the ITS Administrative Controls Section: " Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)" [and

                                        " Post Accident Monitoring (PAM) Report."] The PTLR is more restrictive because it is not currently required.

This change is also described in the description of changes for CTS Section 3/4.4 (PTLR). [The PAM Report is already required per TS 3/4.3.] 03 14 M Shutdown Margin values would be moved to the Core Operating Limits Report (COLR) traveler TSTF 9. In addition, moderator temperature coefficient limits would also be relocated to the COLR. The addition of these specifications to the COLR is considered to be more restrictive. i 03 15 H Refueling boron concentration limits will also be added to the COLR. The addition of these limits to the COLR is considered to be more restrictive. 03 16 A Not applicable to Callaway Plant. See Conversion Comparison Table (Enclosure 3B). 03 17 A Deletes the methodology section references in the COLR. These references are adequately defined by the analytical methods themselves as approved bu the NRC and it is redundant to repeat the information in the ITS. This j i change is consistent with NUREG 1431. _ l l

s

                                                                   -                   =

5 -9( m_l \ 03 19 B--The-ter.T. "unavuw ized" it changed to "inadvcrtent" " t"e X @ " k High Radiation ^re: ;cction. The eicvenn en O L __ 9 5/15/97 I DESCRIPTION OF CHANGES TO CURRENT TS

I

INSERT 3A 9a Q 5.21 03-18-LS-5 The CTS requirement to provide documentation of all challenges to the PORV's or safety valves is deletod. The reporting of pressurizer safety and relief valve l failures and challenges is based on the guidance in NUREG-0694,"TMI-Related Requirements for New Operating Licensees." The guidance of NUREG-0694 states:

                             " Assure that any failure of a PORV or safety valve to close will be reported to the NRC promptly. All challenges to the PORVs or safety valves should be documented in the annual repo:L" NRC Generic Letter 97-02," Revised Contents of the Monthly Operating Report" requests the submittal of less information in the monthly operating report. The generic letter identifies what needs to be reported to support the NRC Performance Indicator Program, and availability and capacity statistics. The generic letter does not specifically identify the need to report challenges to the pressurizer safety and relief valves. This change is consistent l                             with NUREG-1431 as modified by TSTF-258.

INSERT 3A-9b Q b.2-1 CTS 6.12, which provides high radiation area access control alternatives pursuant j to 10 CFR 20.203(c)(2) has been revised as a result of the change to 10 CFR 20 and the guidance in Regulatory Guide 8.3.8. Since the plant requirements rerne:n the i same, except as identified in specific Description of Changes, the change hs ' considered administrative. This change is consistent with NUREG-1431 as modified by TSTF-258. I i

CHANGE NUMBER ((SRC DESCRIPTION [4 5.7.-( l Ed-tea + ent ry 4 - d1 = = ., @cL u u,, 1.5 or as 8.3rr. 2,__ .._,_, ; uJ- m ,- ,, . .. - . ~,, - - -. _-__ ' Il l J Iiu lCIIUbbD LilU IV\b @ p@ l b lUl4 I myus W Illy yp _e.Jas.u

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l j 03 20 LS 3 Not applicable to Callaway Plant. See Conversion 1 Comparison Table (Enclosure 3B). 1 I l 1 i 1 i I 1 l l

   .i t  ,
   ?

t i t 10 5/15/97 DESCRIPTION OF CHANGES TO CURRENT TS

T CONVERSION COMPARISON TABLE - CURRENT TS 6.0 Page 2 of 7 . TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON C0HANCHE PEAK WOLF CREEK CALLAWAY 01-07 Revises Section 6.2.2a. Unit Staff Organization, to reflect No - DCPP is a No CPSES is a Yes - USAR Chapter Yes, moved to FSAR LG the non-licensed operator staffing requirements for a multi-unit plant multi-unit plant 13 ; single unit site. 1he minimum shift crew composition as , described in Table 6.2-1 has been moved to a licensee controlled document. 01-08 Moves the fire brigade requirements to a licensee No Yes - Moved to FSAR Yes, moved to USAR Yes - Moved to the , LG controlled document. These requirements can be found in LA 75/74 FSAR BTP ASB 9.5-1 and their duplication on the ITS is not required. ] ( -eH59- "^t e -A _f rtS e r6 5 B _ 7. A -N/A- -WA-- W -WA-- Q5.'l-(j . 01-10 Adds requirement for three Auxiliary Operators for the two No - Already DCPP Yes No - Wolf Creek is No - Callaway is a 1 M unit sites with both units shutdown or defueled. requirement a single unit plant single unit plant 01-11 For clarity, a note is added to state that one Radiation No - DCPP procedure Yes No - Wolf Creek is No - Callaway is a l A Protection Technician and one Chemistry Technician can and operational a single ur.it plant single unit plant . fulfill the staffing requirements for both units. requirements differ. . 01-12 Deletes the STA qualifications t'ased on use of RG 1.8 No - DCPP has Yes No - Wolf Creek-is No - not in CTS I A Revision 2. additional staff comitted to ' requirements additional requirements 01-13 Adds new statement to accomodate unexpected absences of No - already in CTS Yes No - already in CTS No - already in CTS A on-duty crew member. i 01-14 Remove the requirement for SS and OS to hold a Senior No - DCPP No - Not in CTS No - Wolf Creek has Yes A Reactor Operator license. requirements differ different requirements 02 01 CTS Section [6.6.la] for Reportable Event actions has been Yes No - Deleted from Yes Yes A deleted from the CTS. CTS per Amendnent 50/36 ! 02 02 The " Safety Limit Violation" section is deleted from the Yes Yes Yes Yes ! LS-4 CTS. The NRC reporting requirements are duplicative of 10 CFR 50.72. 10 CFR 50.73 and 10 CFR 50.36.

            .rris u t 3 8- 7.6 V                                  4 Q 5.2-l {                                                                                                                                                                                      e ,, e ,a,
     - . , . . - - - . . , . . - - . . - , - - . . , . . - _...,-,-c

INSERT 38-2a Q 5.2-1 TECH SPEC CHANGE APPLICABILITY NUMBE DESCRIPTION DIABLO CANYON COMMANCHE WOLF CREEK CALLAWAY R .. PEAK 0149 The CTS requirements concerning overtime being in Yes Yes Yes Yes A accordance with the NRC Policy Statement is replaced by referring to administrative procedures for the control of working hours. INSERT 3B-2b Q 5.2-1 TECH SPEC CHANGE l APPLICABILITY NUMBE DESCRIPTION DIABLO CANYON COMMANCHE WOLF CREEK CALLAWAY R PEAK 01-15 A This change revises the CTS to eliminate the title of Yes pygg Yes Yes "ShiftTechnical Advisor (STA)." i _ _ _ . _ _ _ . _ _ _ _ _ _ _ - - _ _ ___m__ _ _ - - _.

i i 4 1 { CONVERSION COMPARISON TABLE - CURRENT TS 6.0 Page 4 of 7 i TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON C0HANCHE PEAK WOLF CREEK CALLAWAY 02-12 This change moves the Emergency Diesel Generator No - DG Failure No - Not in CTS Yes - move to USAR Yes - move to FSAR LG Reliability Program requirement to a licensee controlled reports addressed l document. in Section 3/4.8 l 02-13 Revises Section 6.14 item b to move the requirement that Yes - Requirements Yes - moved to ODCH Yes - requirements Yes - moved to the i LG ODCM (or similar programs and procedures) changes require in plant procedure are in ODCH ODCM review and acceptance by onsite review constittees. AD1.ID2 l 02 14 Per GL 89-01. concentrations of radioactive material Yes No - changes Yes Yes j M releases in liquid effluents to unrestricted area shall included in CTS j conform to 10 times the concentration values in Appendix B. Table 2. Column 2 of 10 CFR 20.1001-20.2401. j 02-15 CTS Section [6.6.1b] contains requirements for the plant Yes No Yes Yes LG review and submittal of a reportable event. This information is to be moved to a licensee controlled document. l 02 16 Change the Diesel fuel Oil Testing Program description for No - DCPP does not No Yes Yes A sampled properties of new fuel oil from "within limits" to currently have a

                   " analyzed" within 30 days following sampling and addition                                                        program description of the new fuel oil to storage tanks. This wording more                                                           in the Admin clearly defines that within 30 days following the initial                                                         section for the new fuel oil sanple, the fuel oil is analyzed to establish                                                        Diesel Fuel Oil that the other pro                 spectriea in i                                                     f           Testing program.

AST W 975-81 ar t. This change is consistent wi Bases for IT SR 3.8.3.3. [3.f'F"- LTl) " 02 17 The React Coolant Pump F1 1 Inspection Program is Yes -- 1 No - See Section No - already in CTS Yes LS-1 revised c nsistent wit The proposed changes -U, Q 5.5 -l , 3/4.4. CN 10 03 tS provide a exception to the examination requirements in Regulatory 1.14. Rev. 1. " Reactor Coolant Pump Flywheel Integrity. N 02-18 Revise the Radioactive Effluent Controls Program dose rate No - already in CTS No - already in CTS Yes Yes

A _ limits to reflect changes to 10 CFR Part 20.g Cait (2: = i c ' ~* *r ::s : - ,ud 'r TA".

                 ~                                                                                                               _

r q s.2.-r _ , .~-_g , u , - Y -

CONVERSION COMPARISON TABLE - CURRENT TS 6.0 Page 5 of 7 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY 02-19 The surveillance interval for verifying that other 'No - addressed in No - addressed in Yes Yes LS-2 properties are with limits for ASTM 2D fuel oil is changed 3/4.8 (CN 3/4.8 (CN from "within 30 days" to "within 31 days" after obtaining a 01-60-LS24) 01-60-LS24) sample. 02-20 Add the provisions of SR 3.0.2 and SR 3.0.3 are applicable No, not in CTS. No, not in CTS. Yes Yes A to the Diesel Fuel Oil Testing Program. This change is based on traveler TSTF-118. 03-01 The method for submitting all reports is revised to be in Yes Yes Yes Yes A accordance with 10 CFR 50.4. 03-02 The requirement to submit a Startup Report is deleted from Yes No - Deleted from Yes Yes A' the CTS. This report required rio staff approval and was CTS per Amenh ent submitted after the fact and is therefore not required to 50/36 ensure safe plant operation. The approved 10 CFR 50 Appendix B. QA Plan, and FSAR startup testing program provides assurance that the affected activities are , adequately performed and that appropriate corrective actions. if required, are taken. 03-03 Revises the annual report section to reflect the new 10 CFR Yes Yes - Except the Yes Yes i A Part 20 requirements and associated reconmended changes Part 20 noted in NRC letter dated July 28. 1995. " Changes to requirements were Technical Specifications Resulting from 10 CFR 20 and removed from the 50.36a Changes." (From Christoper I. Grimes to Owners CTS in Amendment Groups Chairs) - TSTF-152 50/36. The requirement to report specific activity limit Yes Yes Yes Yes 03-04 A violations is deleted. Serious degradation of a fission ! product barrier, among other more serious events are required to be reported by 10 CFR 50.73. This change is a &inistrative in that it only affects reports and dc not affect plant operations. The Annual Radiological Environmental Operating Report is No - DCPP report Yes No No - Callaway 03 05 A revised to change the submittal date. dates to remain the report date to same remain as is in CTS s (gkeh 36-5 ' Q 5.2.-(

8K. INSERT 3B-Sa - Q 5.2-1 TECH SPEC CHANGE l APPLICABILITY , NUMBE DESCRIPTION DIABLO CANYON COMMANCHE WOLF CREEK CALLAWAY R PEAK 02-22 The Radioactive Emuents Controls Program is revised Yes Yes Yes Yes A to include clarification statements denoting that the I provisions of CTS 4.0.2 and 4.0.3 are applicable to these activities. [ r [ i i i I

Page 7 of 7 CONVERSION COMPARISON TABLE - CURRENT TS 6.0 i TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY 03-14 Shutdown Margin values would be moved to COLR per traveler Yes No - Already part Yes Yes M TSTF 9. In addition. moderator temperature coefficient of CTS limits would also be relocated to the COLR. 03-15 Adds refueling boron concentration limits to COLR. Yes Yes No - already in CTS Yes M No Yes No No 03 16 Deletes one of the allowed ECCS evaluation models for CPSES A Unit 2 which is no longer used. Deletes the methodology section references in the COLR. No - References do Yes Yes Yes 03-17 A not exist in DCPP CTS kb 5:0: th: ;prti;.i requir; ;.:t 's- h m.tatica Of b Jio-- -No- -Ves-- E]j

              .A-        -challa g; to thc St er ::fety v Ffer te the El' C. ;;;                                                                                                                                                      r 5' M

{% r wte a;=te.; ne;:re ansert 3s- B A i

                                                                                                             .Ye(                Yer                  Ves-                                                                Yps' 03-19   (The erm " author zed" i cha                            to "inad rt   t-  nt                                                                                                                             l Hi                                              he pr vent        e na er     t      pA                 BJA                 tuh                                                                 PJA           5.2-1]

[ Radi ion A a sec on.

                                                                                       .38. hJot U3d,                                                '

l e ry is discu ed in ectio 1.5 RG _ 7 , The use of a continuous guard is provided as an additional Yes Yes Yes No. maintaining CTS 03-20 LS-3 option for preventing inadvertent entry into high radiation areas that are accessible to individuals.

                                                                                                                                                                                                                                        ,,,,,n,

INSERT 38-8a Q 5.2-1 TECH SPEC CHANGE APPLICAfilLITY NUMBE DESCRIPTION DIABLO CANYON COMMANCHE IVOLF CREEK CALLAWAY R PEAK 03-18 The CTS requirement to provide documentation of all Yes Yes Yes Yes LS-5 challenges to the PORV's or safety valves is deleted. I r s I I r [

ENCLOSURE 4 N0 SIGNIFICANT HAZARDS CONSIDEPATION (NSHC) CONTENTS i I. O rg a n i z a t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 l II. Descri ption of NSHC Eval uations. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 III. Generic No Significant Hazards Considerations A Administrative Changes.............................................. 5 l R - Relocated Technical Speci fications. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 LG Less Restrictive (Moving Information Out of the Techni cal Speci fi cati ons) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 M More Restrictive Requirements....................................... 12 IV. Specific No Significant Hazards Considerations LS d,,.. - LS 1.................................................................... 14 LS 2............................................................ ....... 16 LS 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . No t a ppl i c a bl e LS 4..................................._................................. 18 Ls-5 are f. 4-ap Q 5.2d } ; 1 1 i NO SIGNIFICANT HAZARDS CONSIDERATION 1 5/15/97

1 INSERT 4-a Q 5.2-1 i IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATIONS NSHC LS-5 10 CFR 50.92 EVALUATION FOR TECHNICAL CHANGES THAT IMPOSE LESS RESTRICTIVE REQUIREMENTS WITHIN THE TECHNICAL SPECIFICATIONS The CTS requirement to provide documentation of all challenges to the PORV's or safety valves is deleted. The reporting of pressurizer safety and relief valve failures and challenges is based on the guidance in NUREG-0694,"TMI-Related Requirements for New Operating Licensees." The guidance of NUREG-0694 states: " Assure that any failure of a PORV or safety valve to close will be reported to the NRC promptly. All challenges to the PORVs or safety valves should be documented in the annual report." NRC Generic Letter 97-02," Revised Contents of the Monthly Operating Report" requests the submittal of less information in the monthly operating report. The generic letter identifies what needs to be reported to support the NRC Performance Indicator Program, and availability and capacity statistics. The generic letter does not specifically identify the need to report challenges to the pressurizer safety and relief valves. This change is consistent with NUREG-1431 as modified by TSTF-258. This proposed TS change has been evaluated and it has been determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92(c) as quoted below:

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

1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or I -

3. Involve a significant reduction in a margin of safety."

The following evaluation is provided for the three categories of the significant hazards l consideration standards:

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

The proposed change would not affect the method of operation of plant systems and involves only the deletion of reporting any challenges to the PORVs or safety valves. Reporting of challenges to the PORVs or safety valves has not impact on any accident previously evaluated. Therefore, the proposed change would not result in a significant increase in the probability or , consequences of a previously evaluated accident. 4 3

IV. SPECIFIC NO SIGNIFICANT HAZARDS CONSIDERATION NSHC LS-5 (continued)

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

The assumptions of the accident analyses are unaffected by the proposed change. No new l permutations or event initiators are introduced by the deletion of this reporting requirement. Therefore, this proposed change would not create the possibility of a new or different kind of accident.

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

r l The accident analyses are assumed to be initiated from conditions which are consistent with the j Technical Specifications Limiting Condition for Operation. The proposed change does not affect any LCO. Therefore, there .is no change in the accident analyses and all relevant event acceptance criteria remain valid. Further, the proposed change has no affect on any actual or regulated failure point which is protected by an event acceptance criterion. Because there is no change in any failure point nor in any event acceptance criteria, there is no reduction in a margin of safety. NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Based on the above evaluation,it is concluded that the activities associated with NSHC "LS-5" resulting from the conversion to the improved TS format satisfy the no significant hazards consideration standards of 10 CFR 50.92(c), and accordingly, a no significant hazards consideration finding is justified, l l l l i i

l l Industry Travelers Applicable to CTS Section 6.0/ITS 5.0 ( TRAVELER # STATUS DIFFERENCE # COMMENTS l TSTF 9 Incorporated B-PS NRC Approved i l TSTF 37. Rev.1 Incorporated 5.6 2 DCPP only TSTF 52 Incorporated 5.5 4 'l'I ,($ %M 3.3 g } TSTF 65 Not Incorporated NA Not NRC approved as of ; traveler cut off date. TSTF 106, Not Incorporated NA Retained CTS Rev. 1 TSTF 118 Incorporated 5.5 8 hC. APfrovhlTR 5.o- oof, \ TSTF 119 Not Incorporated NA ($tcir.:d CTS NRt R<icch(TE.5.o.odh

TSTF 12 k) Not Incorporated xA tained CTS jTR 5.o-co( } ffJ1/1 fS8 / [f1npdrporgteg /' // yM'y [$5.2.-41 TSTF 152 Incorporated 5.6 4 {TK 5.o. o%l (J$Tf1fd / / Ingdrp#ated / r' / Jr.7 2'g7 , jQ5.2-/] M p?',tggv.// Incorporated 5.6 5 jn s.o co3 WOG 72 IncorpoNted 5.5 13 x (T_sTF-7. _ f WCC S [ E Incorporated 5.5 14 kQ 6 I~ l Proposed Incorporated 5.2-2.,5.51,5.2-3 Wp nyIni grouf Ac/io Traveler {,1~,[>l.[_f > 55-4 Jteml147( ( l ; Ng

                                                                                                                                                        ~
                                                                                                                                                                                 ~

{TSTF- 2.58 _ i I l w . , . - - - - . - . . . _ _ , . - . . - - - - . - - - - - - - . ~ - - - . - ~ - - - - - - - - - - - -- - - ~ - - - ~ ~ - ~~~- - - - - - ~ ~ f 1 MARK UP 0F WOG STS REV I (NUREG 1431) 5/15/97

Organization 5.2 i 1 5.2 Organization 5.2.2 Unit Staff (continued) rcactor containing fuci and an additional non licensed opcretor shall 5.2 4 bc assigned for cach control roca from which ; rcettor is opcrating 2, _,,ommre

                                        -e       .,.e,        . , , .
                                  ~~?wo unit sitcs with both unit shutdown or dcfucled rcquirc ; total                                    B PS of thr;c non-liccased opcretors for the two unit .

br At icast onc licca;cd Rcactor Opcretor (RO) sh;11 bc prcacnt in the l control ros; whcn fuci is in the rcoctor. In addition, whilc the i unit is in "00 1, 2. 2, or 4, et icast onc liccascd Senior Rcactor 5.2 2 Opcr; tor (SRO) shall bc pres r,t in thc control room. Notiused bser Shift crew composition may be one less than the minimum requirement I of 10 CFP. 50.54(m)(2)(1) and 5.2.2.a and 5.2.2.f for a period of time ED not to exceed 2 hours in order to accommodate unexpected absence of , on duty shift crew members provided immediate action is taken to l restore the shift crew composition to within the minimum requirements. 0 c' id: A bus]tjh]phycineltechhician shall be on site when fuel is in the B reactor. The position may be vacant for not more than 2 hours, in order to provide for unexpected absence, provided immediate action is taken to fill the required s t on pe rs o nnel djer Administrative proced res s all be developed and implemented to limit the working hours of,..s.. :. ' who perform safety related functions 3 d (e.g.,& licensed SenioCiteactor20peratotsI(SR0s), licensed (teictor //AD // O_p_ erat _or.s3(R0s), health physics. tst+ t_e.c.h.n.i.c_ian_s, auxilicry e_quipment. g,t 3 operators, and key maintenance personne ).. g a s.2-t Adcquetc shift coveregc shall bc ;cint;incd without routin- hc;vy use B PS of cycrtiac. The objcctivc shall bc t; havc opcrating pcrsonncl work an [0 cr 12] hour day, nominal 40 hour wcck whilc the unit is opcrati,ig. llaucecr, l,i t,u cec,it t,ut u,i',arcacci graMcas requlrc substantiel ;;sunts of overtiac to b used, or during cxtcadcd periods of shutdown for refueling, major acintenancc, or acjor plant modification, on a tc,T,porary basis the following guidclincs shall bc

                                   ,_ ,,_.._ 1

1. An individual should not bc pcraitted to wc.k acre than 10 hours straight, cxcluding shift turnover ti;^,

i-(continued) MARK UP OF WOG STS REV 1 (NUREG 1431) 5.0 3 5/15/97

_ .... _ _ . _ . _ _ _ __ _ _ -_ __..~._.._ __ _ . . _ _ _ . _ _ . _ _ . _ _ _ . . Organization 5.2 5.2 Organization 5.2.2 Unit Staff (continued)

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                                                                '6.rs. , u'b . .. .., "m-~,,.s- m me, e 'a e, b u . . b^ - - ^~.rs.

s i

                                                                                                                                                                                      . u^b . .. .e "-,~--   s~         ,s. , a
                                                                                                                                                                                                                                '                                                  : B PS" l                    .     .

m, m. ;so c. - ;., s .; . shall hold an SR0 license. _

An Mdud _ ._ _ __ _ to wa1M o penhw sM4 Cec 3 99- ifb Tdedica/ #fvipdr[(ST/ 11 provide advisory technical _ support t; ti '- 'SI.Wt SWM ;? ' Din the areas of thermal i

                                                                                                                                                                                                                                                                                'G  S.2-1 )

hydraulics, reactor engineering, and plant analysis with regard -

                         ~[h s poi &                                                                                                                                                                                                          _

meet the qualifications,_

____,-,_J L.. Au_ n . . n_,2.. eaybb u a a hw ug bsib wvusuus e d a i vi s awI IbJ

         '6
            ~

? l 1 MARK UP OF WOG STS REV 1 (NUREG 1431) 5.0 4 5/15/97 y . _ , . . ~ , ,. - - - . - ~ . , . . -_ - . . - - . - , .

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

I INSERT 5.0-4a Q 5.2 1 The controls shallinclude guidelines on working hours that ensure adequate shift coverage shall l be maintained without routine heavy use of overtime. Any deviation from the above guidelines ; shall be authorized in advance by the { Plant Manager } or the { Plant Manager's } designee, in accordance with approved administrative procedures, and with documentation of the basis for y anting the deviation. Routine deviation from the working hour guidelines shall not be authorized. I I 1 l

Unit Staff Qualifications q 5.3 5.0 ADMINISTRATIVE CONTROLS .5 5,3 Unit Staff Qualifications 1

                                                                                                                                               -             1 Rcviewcr's l4ctc. tiinimum qualifications for acabcrs of thc unit staff shall bc                                                       B PS spccificd by usc cf an overoll qualification statc; cat rcicrencing on #CI St;ndard acccptabic to th liRC staff or by specifying individual position

qualifications. Ocacrally, the first acthod is picfcrobic, howcycr. the second j method is adaptabic to thosc unit staffs requiring spccial qualification statc; cats bccouse cf uniquc org;nizational structurcs. - I 5.3.1 Each member of the unit staff shall meet or exceed the minimum i qualifications of ANSI /.ANS'3:111978EwithJthe2fo110wingiexc.eptipns; 5.3 1

[ Regulatory Guidc 1.0, Rcvision 2,1007, cr acic recent revisions, or MCI B-PS Stand;rd occcptabic to the !!RC staff]. Thc staff not ccvcicd by [ Regulatory Ouide 1.0] shall ;;ct or cxcccd th; minimum qu;lific;tions of [Rcgulations, Regulatory Ouidcs, or MCI St;ndards acccptabic to lRC staff]. l ~' 5;3.T1 Shi ftguperv.isorsE0petating @uperVj sor;sEReactorIOperators . 5.3 1 hhdIShiftHechnical2Adijs6rgsiha))i,neet3orjexceedith;e

gualtUcations a offANS UANS ggi19815 asiendorsed"by; Reg Gui dejl;82Revj sioC2.;wij;h31elsamelexceptionss asicontained li n the]currentire11si#nJtolth' J0perstbr*lli e censing:%aminer S+andardsEMJREG30212ES12021 5.3 ;1'.2 ReIRadiationIProtectj orqManager3 hall 3befa_isuperpisoriwith line responsj bJ1JtyJIorloperationel2healthjphysj esi wholmeets f or exceedslithelqu aMficati onslof'USNRCfRegtil atory1GQi de',"1?8 ;

Septembefl9751;foGaIRat!!ation3rotectionJanager2 The Radj ation7P rot #ctionJjanagetM11Ebeidesi gnated ? by;the: Plant Mahager; 5.3 . 2. For he pi pos<. d4 t o c FR 55. 4, a. la'grud Se m ot Ra n d o r 5. 3 - 2__ o pe va h r C s R.o) uhA h'cc ro d R.ca ch r qu aA r ( R o) a. v e.

                       + h d <. in 6 ad 6vds c % , in c46; b6n bmeeA,'.y N re h re ments                                                       Q 5.2-1 o ( T s s a.I , p4 < 6 <n + % fr.,ncM. m Le n., A t in io cpg so.s 4 (m),

4 s (continued) MARK-UP OF WOG STS REV 1 (NUREG-1431) 5.0 5 5/15/97

__ __ . . . _ _ _ _ _ _ _ _ _ _ _ . _ - __ __ __ _ __ ~__ Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Procram (continued) , b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to 10; times;the I concentration.jalues?ip 10 CO 20, Appendix B. Table 2, Column 2; to 5.5 1 103CFR'20p001"i?072402; R

c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology ,

and parameters in the ODCH: l

d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released fre; ce:h unit to unrestricted areas, conforming to PS 10 CFR 50, Appendix I

e. Octoraination of cumulativ; and projected dosc ;cntributions frea ~

5.5 13 . radioectiv; cfflucnts fer the current calcader quartcr and current l calender y;;r in eccordonce with th; acthodology and par;;;ter; in ' the OOC" et 1; cst cvcry 21 days, {' petegnination3cficumul atiyerdose?contr,1butionsifcom'radioacti ve sff[uients"ifortt}e3 currentic.alendaggua.ttegand (current? cal endarfyearli n pecordance2withjmethoidolggy3ndaarametetsDn2the!ODCHED.et.ermi na. tion Ef'projeMidZdo]e5Contrib0tionsWr;fylicifeti?elefflt!eritsIlti3cc6rdance F1th1the7metho'ddlogyM)Lthe?.00CM1tILea.s_tleyeryg11 days;

f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to -

10 CFR 50, Appendix I; - - L o,n W d 4 cd of 9 Limitations on the dose r resulting rom radioactive material releas d in gaseous effluents to areas beyond the site boundary ryfp t; thc d--- -^-- '_;tsd with 1" C T 20 J escadi t " 5.5 1 Tehlc 2. Column 1, ka gpi Du LL be I W (*d'."'" a +g. i i o t,' ia : (() 5.2-I { l'. For7 noble gasesi M% Nr~ dais)a'doserate'of

                    <      500 mrem /yr;to5the;whole;t iy;and                          Th forfe/(u Q a_ dose rate _of~3000, mrem /.yr;tojthe.skinCand p,

(continued) MARK-UP 0F WOG STS REV 1 (NUREG 1431) 5.0 9 5/15/97

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

l Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Proaram (continued)

2. ForJIodjne 131, hIodine:1337 tritiumTand "for Lall

! radionuclides ~in;particdlate"for@m withihalfflivesigreate QS.2-[ Bjdaysn W $ gnpal%o]a: dor,elratejof 150.0; mrem /yrlto any or.gan. A

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the l site boundary, conforming.to 10 CFR 50, Appendix I:

1. Limitations on the annual and quarterly doses to a member of the public from iodine 131, iodine 133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released fra cech unit to areas beyond the site boundary, conforming PS to 10 CFR 50, ADoendix I and ihe.jon6 W s We 6 6 kry ) ,5 _ ,

j. i Limitat ns on the annual dose or dose commitment to any member of the public ue to relea:es of radioactivity and to radiation from uranium g g,t g } fuel cycle sources, conforming to 40 CFR 190; h 5.5.5 Comoonent Cvelic or Transient Limit ; B PS This program provides controls to track the FSAR. Section E3!9(NT11T

                         " Design 3Transib_iftf3, cyclic and transient occurrences to ensure that components are maintained within the design limits.

5.5.6 "r Strc;;;d C ,nc ;t: Containment Tendon Surveillance Proaram I T5is program provides controls for monitoring any tendon degradation in pic- B PS stice ;cd concrctc centeirac.ts, including effectiveness of its corrosion ! protection medium, to ensure containment structural integrity. The program shall include baseline measurements prior to initial operations. The Tendon Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with the'Eall.away! position.;on]proposedlev1sion~J ofjRegulatory i Guide "195f.dat,ed Apr,il'1979. [Rr uletery Cuidc 1.35, Rcvision 3,1000). t The provisions of SR 3.0.2 and SR 3.0." are applicable to the Tendon Surveillance Program inspection freque. cies. l m

gg.,.~I k l K. 1h 3rM s i nr d S R 3. o.1. d SR 3. o.3 are a.pplicsW h he. R s. G a.t M 4, l EH tvewt %dr b ?<.),m sma gtam 6.Iecy . [ _ y ,Q 5.2-l ,j l. (continued) MARK UP OF WOG STS REV 1 (NUREG 1431) 5.0 10 5/15/97

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.3 Radioactive Effluent Release Reoort (continued) ODCH and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B.1. 5.6.4 Monthlv Ooeratino Reoorts 1 Routine _ reports _of operating statistics and shutdown experience B ! el re 2e 1 ubmitted on a monthly basis no later than the 15th of each month following the ~ $51-l 1 3 calendar month covered by the report. 5.6.5 CORE OPERATING LIMITS REPORT (COLR) I l

a. Core operating limits shall be established prior to each reload '

cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following: _.Inc individual specifications that addic;s cerc cpcrating limits _ must bc rciercaccd here. li Moderator' Temperature.;Coefficientslimitsii n B PS Specification '3:1':3',

2. Shutdown: Bank 3nsertionilimit for' Specification 13.1:5, 3: Cont roe Bang nserti on " Limits 1forJSpeci fi ca ti on Z3.1 6, 4; Axig; Flux Differencellimits forgpecification 3:2:3;

5. HeatTFlux Hot' Channel Jactor, TatZ)' Fa",~ KCZ),' WCZ) and Fa Penalty Factors for Specification 3.2.1,

6. ^

Nuclear Enthalpy? Rise Hot 'ChannerFactor?F.3,"F,n*, and Power Factor Multiplier, 'PF.nimdts for; Specification ~ 3.2.2.

7. Shutdown Margin Limits;forLSpecifications '3.1.1, 3.1.4, 3.1;5, and 3.1.6.

8. Refueling Boron Concentration Limits for Specification. 3.9.1.

(continued) MARK UP OF WOG STS REV 1 (NUREG 1431) 5.0 33 5/15/97

High Radiation Area B In 5 e < t 5. o - 3 '1 5.7

5. 7 - [ Q 5,2-l \

              .0   ADMINISTRATIVE CONTROLS

5. High Radiation Area B

5,7.1 Pursuant to 10 CFR 20, paragraph 20.1601(c), in lieu of th E g.7-1 equirements of 10 CFR 20.1601, each high radiation area as defined i 10 CFR 20, in which the intensity of radiation is > 00 mrem /hr but < 1000 mrem /hr at 30 fem, shall be barricaded and onspicuously poste as a high radiation area and entrance theret shall be contro ed by requiring issuance of a Radiation W k Permit (RWP). Individu s qualified in radiation protection pr cedures (e.g., Health l Ph ics Technicians) or personnel conti uously escorted by B such indivi als may be exempt from the RWP 'ssuance requirement during the pe ormance of their assigned d ies in high radiation areas with exposure r es s 1000 mrem /hr, prov'ded they are otherwise following plant r iation protection p cedures for entry into such high radiation area

    ~~
            ;              Any individual or group f indivi uals permitted to enter such areas                  E    B PS shall be provided with or ccom nied by one or more of the following:

a. A radiation monitorin vice that continuously indicates the radiation dose rate n the area.

b. A radiation mon oring device hat continuously integrates the radiation dos ate in the area nd alarms when a preset integrated d e is received. Ent into such areas with this monitoring vice may be made after he dose rate levels in the area have een established and person 1 are aware of them.

c. An indi idual qualified in radiation prot tion procedures with a radia on dose rate monitoring device, i responsible for prov' ding positive control over the activitie within the area and

           ;                       sh 1 perform periodic radiation surveillance a the frequency
           !                         ecified by the [Radiction Protection "anager]         'alth Physics        R     B Hanagement personnel in the RWP.                                             E i.7-1 5.7.2           n addition to the requirements of Specification 5.7.1, are q with i               radiation levels 2 1000 mrem /hr at 30 cm shall be provided wt h                         >.7-2 locked or continuously guarded doors to prevent unauthorizcd inadvertent entry and the keys shall be maintained under the
,         l                administrative control of the Shift Supervisor /OperatingfSuperviso                   f g    ED

\ (continued ( MARK-UP 0F WOG STS REV 1 (NUREG 1431) 5.0 39 5/15/97

l High Radiation Area i

                                                                                                   ,      5.7 Q 5.2- 1                   g High Radiation Area 5.7.2 x

( ntinued)

                        ,m
                             ;;r. on duty or health physics supervision. Doo shall remain lock , xcept during periods of access by perso                 under an approved RWP thatpl specify the dose rate levels i the immediate work areas and the max .             allowable stay times for ' ividuals in those areas.

In lieu of thiess y time specification the RWP, direct or remote (such as closed'cir it TV cameras) ontinuous surveillance may be made by personnel qualifie in radia n protection procedures to provide positive exposure contro ,ov the activities being performed within the area. 5.7.3 For individual high adiation ar with radiation levels of

                        > 1000 mrem /hr B .7 1

_301cm, accessible o personnel, that are located i within large eas such as reactor con'htinment, where no enclosure o exists fo urposes of locking, or that mr ,et bc cor.tircously B .7 3 I guardc", and where no enclosure can be reason ly constructed around C l _,

"~

th ndividual area, that individual area shall . barricaded and F onspicuously posted, and a flashing light shall be tivated as a warning device. [ l 1 l l l l l c l

t . MARK UP OF WOG STS REV 1 (NUREG-1431) 5.0-40 5/15/97

_ m _ . . _ ..- ._ _ _. . _ _ _ _ . .. . . . _ . _ _ _ _ _ _ _ - . High Radiation Area Inte rf. 5.0-37 5.7 5.0 ADMINISTRATIVE CONTROLS 5."l- 1 5.7 High Radiation Area

                                                                                                                          ,q As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601(a) and (b) of 10 CFR Part 20:

5.7.1 Hiah Radiation Areas with Dose Rates Not Exceedina 1.0 rem / hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation:

a. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.

b. Access to, and activities in, each such area shall be controlled by means of Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area (s) and other appropriate radiation protection equipment and measures. ,

c. Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.

d. Each individual or group entering such an area shall possess:

1. A radiation monitoring device that continuously displays radiation dose rates in the area: or

2. A radiation monitoring device that continuously integrates I the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an l appropriate alarm setpoint, or

3. A radiation monitoring device that continuously. transmits dose rateMormqL37and cumulative doselto a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure with the area, or d

(continued) .

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

High Radiation Area 5.7 5.7 High Area Radiation Area 5.7.1 Hiah Radiation Areas with Dose Rates Not Exceedina 1.0 rem / hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation: (continued) 4, A self reading dosimeter (e.g., pocket ionization chamber , lectronic dosimeter) and, i l (2) Be under the surveillance, as specified in the RWP l or equivalent, while in the area, of an individual l qualified in radiation protection procedures. equipped with a radiation monitoring device that l continuously displays radiation dose rates in the area: who is responsible for controlling personnel l exposure within the area, or (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in l l radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to comunicate with individuals in the area who are covered by such l surveillance.

e. Except for individuals qualified in radiation protection procedures, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them.

5.7.2 Hiah Radiation Areas with Dose Rates Greater than 1.0 rem / hour at 30 l Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation. but less than 500 rads / hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation:

a. Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry, and, in addition: ,

1. All such door and gate keys shall be maintained under the l

administrative control of the [ shift supervisor, radiation l protection manager], or his or her designee. r s (continued) i

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

l High Radiation Area 5.7 l 5.7 High Area Radiation Area 5.7.2 Hiah Radiation Areas with Dose Rates Greater than 1.0 rem / hour at 30 I- Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation. but less than 500 rads / hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation: (continued) i

2. Doors and gates shall remain locked except during periods l of personnel or equipment entry or exit.

b. Access to, and activities-in, each such area shall be controlled by means of an RWP or equivalent that includes specification of radiation dose rates in the immediate work area (s) and other appropriate radiation protection ecuipment and measures.

c. Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.

d. Each individual or group entering such an area shall possess:

1. A radiation monitoring device that continuously integrates the radiation rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or

2. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the area, or

3. A self reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an i,ndividual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area: who is responsible for controlling personnel exposure within the area, or l

l (continued)

High Radiation Area l 5.7 l l 1 5.7 High Area Radiation Area 5.7.2 Hiah Radiation Areas with Dose Rates Greater than 1.0 rem / hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation. but less than 500 rads / hour at 1 Meter from the i Radiation Source or from any Surface Penetrated by the Radiation: 1 (continued) (ii) Be under the surveillance as specified in the RWP or 4 equivalent, while in the area, by means of closed ) circuit television, of personnel quelified in j radiation protection procedures, responsible for l controlling personnel radiation exposure in the I area, and with the means to communicate with and control every individual in the area, or

4. In those cases where options (2) and (3), above, are impractical or determined to be inconsistent with the "As l Low As is Reasonably Achievable" principle, a radiation l monitoring device that continuously displays radiation dose rates in the area.

P'*"" *I # "" . Except for individuals qualified in radiation protection l escuted3b Sd procedure j , entry into such areas shall be made only after dose ' . M .2-Il iddsdh rates in the area have been determined and entry personnel are knowledgeable of them.

f. Such individual areas that are within a larger area M3

                             -@o3tP6Iljk1fs 3,,, Migs.f[MTTORMrJTb Where no enclosure exists S 7~4
     ,su k u N k               for the purpose r# locking and where no enclosure can reasonably g g 2,, q be constructed .ound the individual area need not be controlled cowfs{nvnent'            by a locked door or gate nor continuously guarded, but shall be barricaded, conspicuously posted, and a clearly visible flashing light shall be activated at the area as a warning device.

l 4 (continued) t I

l-JUSTlF1 CATION FOR DIFFERENCES FROM NUREG-1431 NUREG-1431 Section 5.0 This enclosure contains a brief discussion / justification for each marked up technical change to NUREG 1431, Revision 1 (enclosure 5A). 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.

1 CHANGE NUMBER JUSTIFICATION 5.1 1 Not applicable to Callaway Plant. See Conversion Comparison Table (Enclosure 6B). 5.1-2 Revises Section 5.1.1 to maintain CTS. The plant manager approves prior to implementation each proposed test, experiment or modification to systems or equipment that affect nuclear safety and are not addressed in the Final Safety Analysis Report or Technical Specifications. C.' 1 Not applicable to Callaway Plant. See Conversion Comparison Table (Enclosure 6B). 5.2 2 This change deletes Section 5.2.2b, since the requirement for the presence of a reactor operator (RO) or a senior reactor operator (SRO) in the control room is adequately controlled by 10 CFR 50.54(m)(2)(iii) and 50.54(k). The ITS 5.5.2.b requirement that is being deleted will be met through compliance with these regulations and is not required in the TS. Q 5.2-1 This is consistent with traveler TSTF@h , 5.2 3 N$drAd. I n se(t (, A-M gi;;.2-I 5.2-4 Section 5.2.2a describes the unit staff requirements for non-licensed operator staffing for multi-unit sites. This change reflects plant specific requirements for a single unit site and is consistent with the CTS. 5.2 5 Not used. 5.2-6 Thij;x.fian evis se,t46n 52'2f WrefJect .tWCJ.Pf99-the-$tff - teFchni I rije.it (,A "ZA .2-I j- 5.2 7 Not applicable to Callaway Plant. See Conversion Comparison Table l (Enclosure 6B). l ! 5.3 1 This change revises Section 5.3.1 to be consistent with the CTS regardin 1 staff qualifications and training. up S.2.-l}

                                        . t. b

, JUSTIFICATION FOR DIFFERENCES TS 1 5/15/97

1 J INSERT 6A-1a Q 5.2-1 5.2-3 ITS Section 5.2.2d (ISTS 5.2.2e) is revised from specific working hour limits to administrative procedures to control working hours. The proposed changes will

provide reasonable assurance that safe plant operations will not be jeopardized by 1

impaired performance caused by excessive working hours. Specific working hour limitations are not otherwise required to be in the technical specifications under 10 j CFR 50.36(c)(5). Specific controls for working hours of reactor plant staff are described in procedures that require a deliberate decision making process to 2 minimize the potential for impaired personnel performance, and that established ! procedure control processes will provide sufficient controls for changes to that , {_ procedure. These changes are consistent with the recommendation in the April 9, { 1997 letter from C. Grimes to J Davis. Additionally, the ISTS statement " Controls ! shall be included in the procedures such that individual overtime shall be reviewed 3 monthly by the [ Plant Superintendent] or his designee to ensure that excessive ! hours have not been assigned."is being deleted. There is no guidance in Generic j Letter 82-12 that discusses these additional controls. The additional requirement ! to have the Plant Superintendent (or his designee) review hdividual overtime on a ! monthly basis is unnecessary since sufficient administrative controls and policies

exist, as well as the role of the individuals supervisors in supervising personnel l- prevent excessive or abuse of overtime. These changes are consistent with TSTF-

] 258. INSERT 6A-2a Q 5.2-1 ! This change revises ITS Section 5.2.2.f (ISTS Section 5.2.2.g) to describe the current [TS] and to i eliminate the title of " Shift Technical Advisor (STA)." STAS are not used at all plants (the function

may be fulfilled by one of the other on-shift individuals). This Section is revised so that it does not j imply that the STA and the Shift Supervisor must be different individuals. Option 1 of the 1

Commission Policy Statement on Engineering Expertise on Shift is satisfied by assigning an individual with specified educational qualifications to each operating crew as one of the SROs (preferably the Shift Supervisor) required by 10 CFR 50.54(m)(2)(1) to provide the technical expertise on shift. However, the ISTS 5.2.2.g wording of, "the STA shall provide ... support to the l Shift Supervisor...,"is considered to be easily misinterpreted to require separate individuals. f Therefore, the wording is revised so that the STA function may be provided by either a separate Individual or the Individual who also fulfills another role in the shift command structure. This ! change is consistent with TSTF-258. 4 i j' INSERT 6A-2b Q 5.2-1 5.3-2 New paragraph 5.3.2 is added to ensure that there is not misunderstanding when complying with 10 CFR 55.4 requirements. The Definitions in 10 CFR 55.4 state:

                                                     " Actively performing the functions of an operator or senior operator means that an individual has a position on the shift crew that requires the individual to be f                                                  licensed as defined in the facility's technical specifications, and that ...." Placing this paragraph in the ITS meets the 10 CFR 55.4 requirement for defining in the facility's technical specifications the function performed by licenrod individuals per 10 CFR 50.54(m)c Adding this paragraph is consistent with the recommendations in the April 9,1997 letter from C. Grimes to J. Davis and TSTF-258.

CHANGE NUMBER JUSTIFICATION. 5.5 1 This change revises Section 5.5.4, " Radioactive Effluent Controls

                                                               - Program." to' reflect new 10 CFR Part 20 reauirements.fA tra           er is                  ng    ra          to flept's cevTsion M                            k9'2-I sist                  ith        CF    art               .                 " Inst f t 6 A-2c 5.5 2                          This change revises Section 5.5. 3, " Post Accident Sampling," to ensure the capability to obtain and analyze radioactive " iodines" in lieu of
                                                                 " gases." This change is consistent with the CTS and plant practices.

5.5 3 This change revises Section 5.5.8. " Inservice Testing Program " to delete " including applicable supports." This change is consistent with the CTS. 5.5 4 The Containment Leakage Rate Testing Program is added to the ITS consistent with the CTS. The Containment Leakage Rate Testing Program is consistent with traveler TSTF 52. 5.5 5 This change revises Section 5.5.13. " Diesel Fuel Oil Testing Program," to be consistent with the CTS. The details of the method applied to this. test are discussed in the associated SR 3.8.3.3 Bases. [To maintain consistency with the Bases for 3.8.3.3, specific changes to the program description are for sampled properties of new fuel oil from "within limits" to " analyzed" within 31 days following sampling and addition of the fuel oil to storage tanks. This wording more clearly defines that within 31 days following initial new fuel oil sample, the fuel oil is analyzed to establish that the other properties are met.] 5.5 6 Not applicable to Callaway Plant. See Conversion Cor.parison Table

(Enclosure 68).

5.5 7 Revise Section 5.5.12c to delete the reference to 10CFR20, Appendix B. Table 2 Column 2 and add reference to the Standard Review Plan (SRP), Section 15.7.3, Revision 2. 10CFR20, Appendix B. Table 2, Column 2 is > inconsistent with the guidance provided in Section 15.7.3 of the SRP, The first Paragraph of Section 5.5.12 states that "The liquid radwaste quantities shall be determined in accordance with Standard Review Plan, L Section 15.7.3." SRP 15.7.3 states, "The curie content is based on that quantity which would not exceed the concentration limits of 10CFR20, Appendix B. Table II, Column 2." These two tables (Table 2, ! Column 2, and Table II, Column 2) are not equivalent and differ by a i factor of 10. The revision to Section 5.5.12c would refer to SRP, t Section 15.7.3, which contains the methodology for the calculations. This change is consistent with the CTS. 4

5.5 8 A sentence is added to Section 5.5.9 ("The provisions of SR 3.0.2 are

! applicable to the Steam Generator Tube Surveillance Program test j , frequencies") and Section 5.5.13 ("The provisions of SR 3.0.2 and JUSTIFICATION FOR DIFFERENCES TS 2 5/15/97

                      -,n,           ,.. n             . ,                  .                    -                 , . , , , , . , , , - . _ - . . , _

I INSERT 6A-2c Q 5.2-1 After issuance of Generic Letter 89-01,10 CFR 20 was updated. The NRC issued a draft Generic Letter,93-XX, on proposed changes to STS NUREGS based on the new 10 CFR 20. The proposed changes are consistent with the draft generic letter, the April 9,1997 letter from C. Grimes tn J. I Davis (with some exceptions) and traveler TSTF-258. The proposed changes maintain the same overall level of effluent control while retaining the operational flexibility that exists with current TS under the previous 10 CFR 20. These changes are intended to eliminate possible confusion or improper implementation of the revised 10 CFR 20 requirements. l 1 l i I 1 l l I l 1 l 1 1 I

CHANGE i NUMBER JUSTIFICATION l 5.5 15 ITS Section 5.5.11c is being revised consistent with the CTS. The I proposed changes specifies a time interval "of within 31 days after removal" in which a laboratory test of a sample obtained from each of the ESF systems charcoal adsorber must be tested for methyl iodide DenetratioIL F i 5.5-l(o .c n s< f t GA-+cO h lk$ 'I l l 5.6 1 Not applicable to Callaway Plant. See Conversion Comparison Table (Enclosure 6B). g CA -5.o- oo3 5.6 2 This change deletes the EDG Report to reflect the recommendations of GL 94 01, " Removal of Accelerated Testing and Special Reporting Requirements for Emergency Diesel Generators," dated May 31, 1994. l 5.6 3 This change revises the report date in Section 5.6.2, " Annual Radiological Environmental Operating Report" to be consistent with the CTS. 5.64 This change revises Sections 5.6.1 and 5.6.3, " Occupational Radiation Exposure Report" and " Radioactive Effluent Release Report," , i respectively, per NRC letter dated July 28. 1995, " Changes to Technical I Specifications Resulting from 10 CFR 20 and 50.36a Changes" (From i Christopher I, Grimes to Owners Groups Chairs). This is consistent with traveler TSTF 152. 5.6 5 PORVliftsettingsarereferencedinPTLRsectionper*@

                                                                                        --                             ~                              TR 5.o-co3)

h. 6 -(* In te.f t GA- 4b } 5.7.-l l o wit 0C f ttfe

                                                ,Z adia              n sour e.             In k < t G A-4C-Q s.2-(

5.7 2 Thi ha rev es "upauthor' ed" t '"inadvert "i the igh diat' n Ar sect n to flec eN s siti as tat in/

                                               , RG                      tion .5 re rdin hysi                      bar iers or     gh      d1ation Qrg$d8.S as.

is i cons' ent f h tr eler STF- . .Inse tt G A-5.7 3 ITS Section 5.7.3 is being revised consistent with the CTS. The. proposed change deletes the phrase "or that.cannot be continuously guarded" from the ITS to be consistent with the practices of the

                                        - _ _ . current TS which do not have this requirement'for high' radiation areas.

l The use of barricades and flashing lightc, is adequate protection for individual high radiation areas to limi. personnel accessibility. ! / 5.l-4 Inkv 6 (. A - 4 e - d2 5.2-I ( 3~ i

                .                                               gD                                     C A-5.o- oo

( JUSTIFICATION FOR DIFFERENCES TS 4 5/15/97

l l l l lNSERT 6A4a Q 5.2-1 5.5-16 The Radioactive Effluents Controls Program is revised to include clarification statements denoting that the provisions of SR 3.0.2 and SR 3.0.3 are applicable to these activities. These statements of applicability clarify the allowance for surveillance frequency extensions and allowance to perform missed surviellances. Generic Letter 89-01," implementation of Programmatic Controls for Radiological Effluent Technical Specifications and the Relocation of Details of RETS to the Offsite Dose Calculation Manual or the Process Control Program" allowed licensees to relocate the Radiological Effluent Technical Specifications and establish the Radioactive Effluents Control Program in the Administrative Controls Section of the Technical Specifications. This change effectively implements the CTS requirements that were relocated per Generic Letter 98-01. Since this change adopts previous CTS requirements,it is considered a change of presentation method only. This change is consistent with TSTF 258. INSERT 6A4b Q 5.2-1 5.6-6 The ITS requirement to provide documentation of all challenges to the pressurizer power operated relief valves or pressurizer safety valves is deleted. The reporting of pressurizer safety and relief valve failures and challenges is based on the guidance in NUREG-0694,"TMI-Related Requirements for New 1 Operating Licensees." The guidance of NUREG-0694 states:" Assure that any i failure of a PORV or safety valve to close will be reported to the NRC promptly. All challenges to the PORVs or safety valves should be documented in the annual report." NRC Generic Letter 97-02," Revised Contents of the Monthly Operating Report" requesta the submittal of less information in the monthly operating report. The generic letter identifies what needs to be reported to support the NRC Performance Indicator Program, and availability and capacity statistics. The generic letter does not specifically identify the need to report challenges to the pressurizer safety and relief valves. This change is consistent with TSTF-258. l l lNSERT 6A4c Q 5.2-1 Section 5.7 is revised in accordance with 10 CFR 20.1601(c) and updates the acceptable alternate controls to those given in 10 CFR 20.1601. These changes are consistent with the draft Generic Letter (93-XX) on proposed changes to STS NUREGs based on the new 10 CFR 20 and the letter from C. Grimes, NRC, to J. Davis, NEl dated April 9,1997. This change is consistent with TSTF 258 and encompasses the NRC comments on 6/11/98. Additional technical changes made to Section 5.7 are identified and justified. INSERT 6A4d Q 5.2-1 ITS 5.7.2.e is revised consistent with CTS 6.12 that allows any individual or group of individuals to enter a high-high radiation area (dose rates greater than 1.0 rem /nour at 30 cm) accompanied by an individual qualified in radiation protection procedures with a radiation dose rate monitoring device. The qualified individual is responsible for providing positive control and shall perform periodic radiation surveillances at the frequency specified in the RWP. The CTS requirements allow the qualified individual to enter a lockea high radiation area with plant workers without first having to enter the area to determine dose rates and then exit the area to provide dose rate information to the plant workers and then reenter the area. This flexibility is in keeping with the

l l L "As Low As Reasonably Achievable" principle while maintaing appropriate radiation worker

practices.

INSERT 6A-4e Q 5.2-1 ITS 5.7.2.f is revised consistent with CTS 6.12 to delete the phrase "that is controlled as a high radiation area". The proposed change would preclude having to post an area around the high-high radiation area as a high radiation area when the area may not meet the definition of a high radiation area. I i i i l l l l t I

CONVERSION COMPARISON TABLE FOR DIFFERENCES FROH NUREG-1431. SECTION 5.0 Page 1 of 4' TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY 5.1-1 Revises Section 5.1.1- . maintain Wolf Creek's current No No Yes No technical specificatioics (CTS). The Plant Manager does not currently approve prior to i glementation each proposed test. experiment or modification to systems or equipment that affect nuclear safety. " 5.1-2 Revises Section 5.1.1 to maintain Callaway plant's CTS that No No No Yes j the plant manager approves prior to iglementation each proposed test. experiment or modification to systems or equipment that affect nuclear safety and are not addressed in the Final Safety Analysis Report or Technical ' Specifications. 5.2-1 Revises Section 5.2.2.a to reflect the current TS. This Yes Yes No - Wolf Creek is No - Callaway is a change clarifies the application of the unit staff a single unit single unit plant. provisions to both units. plant. S.2-2 The requirement for the presence of a RO or a SRO in the Yes Yes Yes Yes control room may be deleted from the ITS consistent with the NRC Policy Statement because this requirement is adequately controlled by 10 CFR 50.54(m)(2)(111). (2-3 JigforSed Ir\Se r t GS-la EK NK W Er'NG S.7.-( } 5.2-4 Section 5.2.2.a describes the unit staff requirements for No - DCPP is a No - CPSES is a Yes Yes non-licensed operator staffing for multi-unit sites. This multi-unit plant. multi-unit plant. change reflects plant specific requirements for a single unit site and is consistent with the CTS. 5.2-5 Not used. N/A N/A N/A N/A 5.2-6 Revises sectiongeflect the CTS . Yes Yes - LA 50/36 Yes Yes oAh . el.mlu4 e W t'alLt. o $ sti$.t moved text to FSAR C

                            ;    T cWEt c al A A v'ts of C.5TA)

I - to fill STA position. Q 5.L.l } r'nanernetnas r nun Antenu Taos e tu rntr' .1 A"31 g/1R/07

INSERT 6B-1a Q 5.2-1 TECH SPEC CHANGE I APPLICABILITY NUMBE DEOCRIPTION DIABLO CANYON COMMANCHE WOLF CREEK CALLAWAY R PEAK 5.2-3 ITS Section 5.2.2d (ISTS 5.2.2e) is revised from specific Yes Yes Yes Yes working hour limits to administrative procedures to control working hcurs. f

CONVERSION CONPARISON TABLE FOR DIFFERENCES ~FROM NUREG-1431. SECTION'5.0 Page 2 of 4 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO "ANYON COMANCHE PEAK WOLF CREEK CALLAWAY 5.2-7 Revises 5.2.2c for Comanche Peak to add note that a single No Yes No No Radiation Protection Technician and a single Chemistry Technician may fulfill the requirements for both units. 5.3-1 Revises Section 5.3.1 to be consistent with CTS regarding Yes - LA 43/42 Yes Yes Yes plant staff qualifications and training. 5.5-1 Revises Section 5.5.4." Radioactive Effluent Controls Yes Yes - LA 42/28 Yes Yes , Program." to reflect new 10 CFR Part 20 requirements. LA 85/84 (c/ns/stp6tpitya/prppo#dtydveltrg f Q 5,7_( (, 5.5-2 Revises Section 5.5.3. " Post Accident Sanpling." to ensure Yes Yes- Yes .Yes the capability to obtain and analyze radioactive " iodines" in lieu of " gases." This change is consistent with t N CTS and plant practices. ; 5.5-3 Revises Section 5.5.8. " Inservice Testing Program." 10 Yes Yes Yes Yes delete " including applicable supports." This change is consistent with the CTS. 5.5 4 The Containment Leakage Rate Testing Program is added to Yes Yes - LAR 96-002 Yes Yes the ITS consistent with the CTS. The Containment Leakage LA 110/109 Rate Testing Program is consistent with TS generic change item TSTF-52. 5.5-5 Revises Section 5.5.13. " Diesel Fuel Oil Testing Program." Yes Yes Yes Yes to be consistent with the CTS. The detLils of the method applied to this test are discussed in the associated SR 3.8.3.3 Bases. t - 5.5-6 Additional programs are added as additional sections to the Yes Yes No No ITS. I 5.5 7 Delete the reference to 10CFR20. Appendix B. Table 2 No No No Yes Column 2 for Callaway Plant and add reference.to the , Standard Review Plan. section 15.7.3 Rev. 2. Add the provisions of SR 3.0.2 as applicable to Section Yes Yes Yes Yes 5.5-8 5.5.9 and SR 3.0.2 and 3.0.3 as applicable to Section 5.5-13. This change is based on traveler TSTF-118. 5.3 -?_ I AG rt 4 8 -2.Q ,' G 5. 7,( g pgg fn7 e.--,a.. e n. .n . n v e n. . v a n e- . .n e-n +ss,

INSERT 6B-2a TECH SPEC CHANGE I APPLICABILITY NUMBE DESCRIPTION DIABLO CANYON COMMANCHE WOLF CREEK CALLAWAY R PEAK , 5.3-2 New paragraph 5.3.2 is added to ensure that there is Yes Yes Yes Yes not misunderstanding when complying with 10 CFR 55.4 requirements. i t t 1

CONVERSI0M COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431, SECTION 5.0 Page 3 of 4 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON , COMANCHE PEAK WOLF CREEK CALLAWAY 5.5-9 Rev1=es Section 5.5.12c to clarify for Diablo Canyon that Yes No No No temporary outdoor liquid radwaste tanks are covered under the surveillance program. ihis is consistent with current plant practices. l 5.5-10 This change lists the tanks that the surveillance program No - DCPP CLB does No - Not part of Yes Yes in Section 5.5.12c is applicable to as is in the CTS. This not list tanks. CTS. maintaining change is a plant specific requirement :onsistent with the ITS wording. CTS. 5.5-11 The documents referenced for the testing frequency for the No - See CN 5.5-12 Yes Yes Yes Ventilation Filter Testing Program (VFTP) do not provide frequencies for combined pressure drop tests or the heater power rating tests. The current TS frequency is added for these tests. 5.5 12 The referenced frequencies for the tests listed in the Yes No No No Ventilation Filter Testing Program (VFTP) were evaluated as part of the 24 month fuel cycle program for DCPP (see LAR 96-09). _ 5.5-13 Revises Radioactive Effluent Controls Program dose Yes Yes Yes Yes projections to meet original intent of TS prior to 1 implementation of GL 89 01. (WOG-72) [ST'F--73g O5.54\ 5.5-14 Section 5.5.7 is being revised consistent th-imd [ ]. tion Yes j

                                                                                                                         /   Yes                  Yes          Yes The proposed changes to Section 5.5.7 provide to the examination requirements in Regulatory Guide 1.14 Revision 1. " Reactor Coolant Pump Flywheel Integrity."

5.5-15 This chang

provides a time interval of within 31 days No No No Yes following removal in which a laboratory test of a sample obtained from the charcoal adsorber must be tested. This change is consistent with the Callaway CTS.

5.6 1 Revises Section 5.6.4. " Monthly Operating Report." for No Yes No No Coraanche Peak tc reflect a revised submittal date. This change is consistent with the CTS. 5 . S - Uo Insert GG-4cQ [cqs.z-1( comwrx,em co-mee no, c . m,,m.uu C1muO l ca-sm-od s,,s,,,

INSERT 6B-40 Q 5.2-1 TECH SPEC CHANGE APPLICABillTY , NUMBER DESCRIPTION DIABLO CANYON - COMMANCHE WOLF CREEK CALLAWAY . PEAK 5.5-16 The Radioactive Effluents Controls Program is revised to Yes Yes Yes Yes include clarification statements de snoting that the ' provisions of SR 3.0.2 and SR 3.0.3 are applicable to . i -' these activities. 1 I i i i

                                                                                                                                                           .I
                                                                                                                                                           'f r

i i b

CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431. SECTION 5.0 Page 4 of 4 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK Call >MAY , t 5.6-2 D31etes the EDG Report to reflect the recomendations of Yes No - not in CTS Yes Yes GL 94 01 " Removal of Accelerated Testing and Special Reporting Requirements for Emergency Diesel Generators." dated May 31. 1994. 5.6 3 Revises report date in ITS 5.6.2. " Annual Radiological Yes - Consistent Yes - See LA 42/48 Yes Yes Environmental Operating Report" to be consistent with the with CTS and , CTS. LA 78/77. 5.6-4 Revises Sections 5.6.1 and 5.6.3. Occupational Radiation Yes Yes Yes Yes t Exposure Report" and " Radioactive Effluent Release Report." respectively, per NRC letter dated July 28, 1995. " Changes to Technical Specifications Resulting from 10 CFR 20 and 50.36a Changes" (From Christopher I. Grimes to Owners Groups Chairs). (TSTF-152) r-5.6-5 [ ] PORY lift se tinm are r=ferenced in PTLR section pe Yes ' TR-5.o-co3) Yes Yes Yes Q C7 k. 1. STF- 7.3M 5.7-1 [RevisesHighRadiationAreatoincorporateconsistent Yes g g,[ Yes Yes Yes , InSeit 66-sb s

                   , changes with 10 CFR 20.1601.                                                                                                                                            <

5.7-2 [Ch ges unaut reed' o ~1 adver ent" t Hig -Ves---- #s--- 6 tes-- Ra lati n Are sectio to r lect the 's si on as I s ated in RG .38. 5 ction .5 r gardi ph ic barr ers I Q 5.2.-(j f r Hijah Rad ation eas. This is to iste t w th tr veler ( TF-1$7 7 I f\5c (f: 66-Scj

                    -         ~

5.7-3 This change deletes the phrase "or that cannot be No No No Yes continuously guarded" from the ITS for Callaway to make them consistent with the CTS. 5.7 -h f A fe rf. G IQ 5.7. - [ 5.c-6 .rnu,t66-SA M kSL I( C AtturDC T Att ('OtJD A D T cent TAnse ste entr= 1 A11 estesn'

INSERT 6B-Sa Q 5.2-1 TECH SPEC CHANGE APPLICABILITY NUMBE DESCRIPTION DIABLO CANYON COMMANCHE WOLF CREEK CALLAWAY R PEAK 5.6-6 The ITS requirement to provide documentation of all Yes Yes Yes Yes challeriges to the pressurizer power operated relief valves or pressurizer safety valves is deleted. INSERT 68-5b Q 5.2-1 TECH SPEC CHANGE I APPLICABILITY NUMBE DESCRIPTION DIABLO CANYON COMMANCHE WOLF CREEK CALLAWAY R PEAK 5.7-1 Section 5.7 is revised in accordance with 10 CFR Yes Yes Yes Yes 20.1601(c) and updates the acceptable alternate controls to those given in 10 CFR 20.1601. INSERT 6B-Sc Q 5.2-1 TECH SPEC CHANGE I APPLICABILITY NUMBE DESCRIPTION DIABLO CANYON COMMANCHE WOLF CREEK CALLAWAY ; R PEAK 5.7-2 ITS 5.7.2.e is revised consistent with CTS 6.12 that F f Yes g ygg allows any individual or group of individuals to enter a . high-high radiation area (dose rates greater than 1.0 remlhour at 30 cm) accompanied by an indivdual qualified in radiation protection procedures with a . radiation dose r9te monitoring device. INSERT 6B-5d Q 5.2-1

TECH SPEC CHANGE I APPLICABILITY NUMBE DESCRIPTION DIABLO CANYON COMMANCHE WOLF CREEK CALLAWAY : R PEAK 5.7-4 ITS 5.7.2.f is revised consistent with CTS 6.12 to delete A yg k yg Yes 7 the phrase "that is controlled as a high radiation area". l The proposed change would preclude having to post ; an area around the high-high radiation area as a high radiation area when the area may not meet the l definition of a high radiation area. i i i o L

ADDITIONAL INFORMATION COVER SHEET l l ADDITIONAL INFORMATION NO: O 5.3-1 APPLICABILITY: DC, WC, CA REQUEST: ITS 5.3.1 (Wolf Creek, Callaway and Diablo Canyon) Comment: Part 55 of Title 10 of the Code of Federal Regulations was revised in March 1987 to establish upgraded requirements for licensed reactor operators. NRC Regulatory Guide (RG) 1.8, Revision 2, April 1987, describes methods acceptable to the staff for complying with the revised rule. The Statements of Consideration for the Part . 55 rule change state that, "Those facility licensees that have made a commitment that is ! less than that required by the new rules must conform to the new rules automatically." The staff is concerned some facilities continue to have technical specifications that reference older industry standards that may not fully meet the revised requirements of l 10 CFR Part 55. I The staff previously considered that the standards applied through the industry's accreditation process were equivalent to the guidance contained in RG 1.8, Revision 2. However, the staff has recently found that current INPO guidance in this area is very general; only advising licensees to follow regulatory requirements. In RG 1.8, Revision 2, the NRC staff endorses, with conditions, certain parts of industry standard ANSI /ANS-3.1-1981 as an acceptable approach for complying with the qualification and training requirements of 10 CFR Parts 50 and 55. This endorsement applies to the positions identified as shift supervisor, senior operator, licensed operator, shift technical advisor, ) and radiation protection manager. For positions other than those identified, the RG

finds acceptable the approach provided in ANSI N18.1-1971. l For Callaway, the ITS proposes to adopt the CTS which adopts ANSI /ANS 3.1-1978 for the unit staff (besides SROs, ROs and STAS) and RG 1.8, September 1975 for the radiation protection manager. For Wolf Creek, the ITS proposes to adopt the CTS which adopts ANSI /ANS 3.1-1978 for the unit staff (besides SROs and ROs) and RG 1.8, September 1975 for the radiation protection manager. For Diablo Canyon, the ITS proposes to adopt the CTS which adopts ANSI /ANS 3.1-1978 for the unit staff (besides the radiation protection manager) though it does makes a reforence to ROs and SROs having to meet the minimum qualifications of Part 55. Please describe how your commitment to an ANSI standard other than that endorsed by NRC RG 1.8, Revision 2 currently meets the requirements of 10 CFR Part 55, as discussed in the Statements of Consideration for the rule change and would meet those requirements with the ITS as proposed. FLOG RESPONSE: Callaway, Wolf Creek, and Diablo Canyon believe that this question is outside the scope of the ITS conversion process because it is a generic industry question. The NRC's question regarding compliance with 10 CFR Part 55 should be discussed on a generic basis. Therefore, it is proposed that this issue be resolved separate from the ITS conversion and the submitted ITS 5.3.1 remain as is which is consistent with the CTS.

l l l ATTACHED PAGES: None. l l 1 l l l l

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 5.5-1 APPLICABILITY: DC,CA REQUEST: Change 2-17 LS1 (Callaway, Diablo Canyon) Comment: WOG-85 has not yet become a TSTF. Use current ITS, FLOG RESPONSE: See response to Comment Number 5.5-2. ATTACHED PAGES: Encl. 2 6-19b Encl. 3A 5 Encl. 3B 4 i I

FRon HRC ADMINISTRATIVE CONTROLS j _ j PROCEDURES AND PROGRAMS (Continued) l

a. Evalonive Gas and Storaan Tank Radioactivity Monitorina proaram l (Conttnued) f The program shall include:

! 1. The limits for concentrations of hydrogen and oxygen in the WASTE CAS HOLDUP SYSTEM and a surveillance program to ensure

the limits are maintained.

any mbe4 in 46 ume*k+d 4

2. A surveillance progr to ensure that the quantity of i radioactivity con ned in each gas storage tank is less

! than the amount at would result in a whole body exposure i of 10.5 rea to : "E"*ER OF E ""!L!C et th :: = :t !!TE-

                               -90WNBARV-in the event of an uncontrolled release of the y"        !

i' tanks' contents, consistent with Branch Technical Position ! ETSB 11-5, " Postulated Radioactive Releases due to Waste Gas System Leak or Failure,' in NUREG-0800, July 1981. j i 3. A surveillance program to ensure that the quantity of i radioactivity contained in the following outdoor liquid l l radwaste tanks, that are not surrounded by liners, dikes, or i

walls capable of holding the tanks' contents and that do not ,

i c have tank overflows and surrounding area drains connected to l the liquid radmasta system, is less than f.he amount that i ! would result in concentrations less than the limits of 10 . CFR Part 20.1 -20.602, Appendix B (redesignated at ! 56FR23391,May21,1991) at the nearest potable water supply 1 and the nearest surface water supply in en UNRESTRICTED AREA, in the event of an uncontrolled release of the . tanks' {

contents

! a. Reactor Makeup Water Storage Tank, ( b. Refueling Water Storage Tank.

c. Condensate Storage Tank, and i d. Outside temporary tanks, excluding deelneralizer i vessels and the liner being used to solidify I radioactive waste.

} The provisions of Specifications 4.0.2 and 4.0.2 are

applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance freqw ncies.

b. Raaetor Coolant pa=a Flywheel Insnaction Prearam Each reactor coolant pump flywheel shall be inspected Q 5.5 -[3 recosumendationsofRegulatoryPositionC.4.bofRegulaNrthe ry Guide 1.14, Revision 1, dated August 1975.TInse rt 3A) 02-n -L5 - 1

c. Containment Tendon Surveillance Proaram This program provides controls for monitoring tendon performance, including the effectiveness of the tendon corrosion protection

medium to ensure containment structural integrity. The program shall lnclude baseline measurements prior to ' nitial plant CALLAWAY - UNIT 1 6-19b Amendment No. Ma,112

l l .. 1(\Se{b 3 A Q 5.5 In lieu of Position C.4.b(1) and C.4.b(2), a qualified in-place UT examination over th: volume from the inner bore of the flywheel to the circle one-half of the outer radius or a surface ! : examination (MT and/or PT) of exposed surfaces of the removed flywheels may be conducted at -

- approximately 10 year intervals coinciding with the Inservice Inspection schedule as required by '

) ASME Section XI. l i j 1 i t I

CHANGE NUMBER RS.tiG DIS.Cf@Il0B 02 10 M Not applicable to Callaway Plant. See Conversion 4 Comparison Table (Enclosure 3B). 02 11 M New program requirements. " Safety Function Determination Pragram" and " Bases Control Program" would be added, consistent with NUREG 1431. Although these new programs reflect current plant practice, delineating them in the ITS would be more restrictive. 02 12 LG Move the Emergency Diesel Generator Reliability Program requirement to a licensee controlled document. Moving 4 this program is consistent with NUREG 1431. Rev. 1. 02 13 LG Revises Section 6.14 item b to move the requirement that ! ODCM (or similar programs and procedures) changes require review and acceptance by onsite review committees to the ODCM. The onsite review of 00CM changes is currently required per [ procedures). This change is consistent with NUREG 1431. 02 14 M Per GL 89 01, concentrations of radioactive material releases in liquid effluents to unrestricted area shall conform to 10 times the concentration values in Appendix B. Table 2. Column 2 of 10 CFR 20.100120.2401. {Prepc;cd trncici pcr. die $rMert 3 A-5Q Lgg.2[} 02 15 LG CTS Section [6.6.lb] contains requirements for the plant review and submittal of a reportable event. This information is to be moved to a licensee controlled document. Given that these reviews and submittal of results are required following the event without a specified completion time, the requirements are not necessary. The moving of this information maintains consistency with NUREG-1431, Rev. 1. 02-16 A To maintain consistency with the Bases for TS 3/4.8.1. change the Diesel Fuel Oil Testing Program description for sampled properties of new fuel oil from "within limits" to

                                  " analyzed" within 30 days following sampling and addition of the fuel oil to storage tanks. This wording more clearly defines that within 30 days following the initial new fuel oil sample, the fuel oil is analyzed to establish that the other properties specified in Table 1 of ASTM D975-81 are met. This change is consistent with the Bases for SR 3.8.3.3 of NUREG 1431.

TST F 1.'5'1 2-17 LS 1 The " Reactor Coolant Pump F heel Inspection Program" is revised consistent with WOC-CT. The proposed changes to provide an exception to the examination requirements in Regulatory Guide 1.14. Rev.1. "Reactcr Coolant Pump (Q55-l} DESCRIPTION OF CHANGES TO CURRENT TS 5 5/15/97

CONVERSION COMPARISON TABLE - CURRENT IS 6.0 Page 4 of 7 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY t~ 02-12 This change moves the Emergency Diesel Gencrator Ma - DG Failure No - Not in CTS Yes - move to USAR Yes - move to FSAR LG Reliability Program requirement to a licensee controlled reports addressed document. in Section 3/4.8 0? 13 Revises Section 6.14 item b to move the requirement that Yes - Requirements Yes - moved to ODCH Yes - requirements Yes moved to the LG 00CH (or similar programs and procedures) changes require in plant procedure are in 00CH 00CH review and acceptance by onsite review comittees. ADI.ID2 02-14 Per GL 89-01. concentrations of radioactive material Yes No - changes Yes Yes H releases in liquid effluents to unrestricted area shall included in CTS conform to 10 times the concentration values in Appendix B. Table 2. Column 2 of 10 CFR 20.100120.2401. 02-15 CTS Section [6.6.1b] contains requirements for the plant Yes No Yes Yes LG review and submittal of a reportable event. This information is to be moved to a licensee controlled document. 02 16 Change the Diesel Fuel Oil Testing Program description for No - DCPP does not No Yes Yes A sampled properties of new fuel oil from "within limits" to currently have a

                 " analyzed" within 30 days following sampling and addition                                                                                                                                                program description of the new fuel oil to storage tanks. This wording more                                                                                                                                                   in the Admin clearly defines that within 30 days following the initial                                                                                                                                                 section for the new fuel oil sample. the fuel oil is analyzed to establish                                                                                                                                                Diesel fuel Oil that the other pro specified ir. Table-1 4 f                 Testing program.

ASTHD975-81 ar t. This change is consistent wit .e Bases for IT SR 3.8.3.3. [5'lF - U b 02-17 The Reactc Coolant Pump F1 heel Inspection Program is , Yes 1 No - See Section No - already in CTS Yes LS.1 revised c nsistent with The proposed changes - Q 5.5-l u 3/4.4. CN 10 03.LS provide a exception to the examination requirements in Regulatory a 1.14. Rev.1. " Reactor Coolant Punp Flywheel Integrity. 02-18 Revise the Radioactive Ef fluent Controls Program dose rate No - already in CTS No - already in CTS Yes Yes A limits to reflect changes to 10 CFR Part 20 ft- {Cencric Let+c =d prcpo rd t m eier. ~ w - i Q 5.2.-(

l 1 1 1 ADDITIONAL INFORMATION COVER SHEET l l ADDITIONAL INFORMATION NO: Q 5.5-2 APPLICABILITY: DC, CP, WC, CA l REQUEST: Difference 5.5-14 I Comment: WOG-85 has not yet become a TSTF. Use current ITS. i FLOG RESPONSE: WOG-85 has been approved by the TSTF and is designated as TSTF-237. This traveler has been submitted to the NRC and is under review. The proposed wording in TSTF-237 was modified from WOG-85 and these modifications have been incorporated l into the ITS. The FLOG continues to pursue the changes proposed by this traveler. For Wolf Creek, this change was approved by the NRC in Amendment No.106, dated June 24,1997. Therefore, the wording in ITS 5.5.7 is consistent with Amendment 106. ATTACHED PAGES: Encl. 5A Traveler status page, 5.0-11 I Encl.6A 3 Encl. 6B 3 \ l \ l l l l l

Industry Travelers Applicable to CTS Section 6.0/ITS 5.0 I TRAVELER # STATUS DIFFERENCE & CQMMENTS TSTF 9 Incorporated B-PS NRC Approved TSTF 37, Rev.1 Incorporated 5.6 2 DCPP only TSTF 52 Incorporated 5.5-4 'f[7 ', d[ 3 3.3.._[] l TSTF 65 Not Incorporated NA Not NRC approved as of traveler cut off date. 1 TSTF 106, Not Incorporated NA Retained CTS I Rev. 1 TSTF 118 Incorporated 5.5 8 h (Trovhf7R Ro-ooG:, T TSTF 119 Not Incorporated NA (35t;in;d-CTS IJRCR<icch(TE5.o oof,3'

TSTF 12 k ) Not Incorporated %3 tained CTS I R T 5 o-cot T hgISJ1/1 f / /[/Inp6rporfteg [ // y' , [$ 5.2-() I TSTF 152 Incorporated 5.6-4 {TR 5 o. oof.1 (7$TY lyf f f Ing6rogfated / / / y.7[g7 lQ52-I} yc/p?',t3gv./g Incorporated 5.6 5 ]p s.o co3

   #                                WOGg                  Incorporated                     :i.5 13                                                                       I (TSTF-1 L NE SS.)                      Incorporated 5.5 14                         kQ 6 5- l Proposed                Incorporated              5. 2- Z., 5. 5 1, i. 'l- 3           Wya rff ni grouf Ac/io Traveler                                                   5

{,1~,t 3,[f ,5 5-4 Jteml147( ( l TSTF- 258 _ -

                                                                                                      '                            ~
                                                                                                                                                  -i      g l

MARK UP 0F WOG STS REV 1 (NUREG 1431) 5/15/97

1 Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.7 Reactor Coolant Pumo Flywheel Insoection Procram This program shall provide for the inspection of each reactor coolant pump i flywheel per the recommendations of Regulatory. Position c.4.b of Regulatory j Guide 1.14. Revisior, ust 1975. l In]ieuZoffPos.jtion1 E4jb(1)' andl.,C'.'4:b(2)J',';;d;tla.; qualified uiniplac_e,yT 5.5 14 examination ~ove~r?t Wdl ume 'f rom ".t he ?.i nne r" b~o reTof :t he7flywhesl 'tolthe

cec 1Mine5.a[.[hNodt'erEdiUs3MdMtMilr; face [lejaniipit[o'nj(HT @55 2} W and/orRT)loffexpose.disuffac.eQbffthelemovedi'99h_ee_1 sarMJ3Ibelconduct_eQat jyeatQntervals]coincidingpithTtheyns' rViceEInspection3chedule'.a's e

           required *bV7ASME Secti6n!XII gpg mw.ly ~~ ~                                                   NY 5.5.8 Inservice Testina Procram This program provides controls for inservice testing of ASME Code Class 1,
           .' . and 3 components including yri mumm -ryu m . The program shall                                    .5.5-3 i include the following:                                                                                       l

a. Testing frequencies specified in Section XI of the ASHE Boiler and Pressure Vessel Code and applicable Addenda as follows:

ASME Boiler and Pressure Vessel Code and applicable Addenda te minology for Required Frequencies inservice testing for performing inservice i l activities testina activities Weekly At least once per 7 days l Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies for performing inservice testing activities:

(continued) HARK UP 0F WOG STS REV 1 (NUREG 1431) 5.0 11 5/15/97

CHANGE NUMBER JUSTIFICATIQB

 )                SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test frequencies") to provide consistency with current application of these requirements. This is consistent with the use of current TS and alleviates potential confusion in the program descriptions. This change is consistent with traveler TSTF-118.

5.5 9 Not applicable to Callaway Plant. See Conversion Comparison Table (Enclosure 6B). 5.5 10 Section 5.5.12c specifics a surveillance program to ensure that the quantity of radioactivity contained in all outside liquid radwaste tanks that are not wrrounded by liners, dikes, or walls is less than the predetermined quantities. This change lists the tanks that the surveillance program is applicable to as is in the current licensing basis. This change is a plant specific requirement consistent with the CTS. 5.5 11 The documents referenced for the testing frequency for the Ventilation Filter Testing Program (VFTP) do not provide frequencies for combined pressure drop tests or the heater power rating test. The current TS frequency is added for these tests. 1 5.5 12 Not applicable to Callaway Plant. See Conversion Comparison Table je (Enclosure 68). 5.5-13 This change revises Radioactive Effluent Controls Program dose projections to meet original intent of TS prior to implementation of GL 89-01. GL 89 01 provided the wording for the STS (Section 5 5.4.e) which combined the requirements for cumulative and projected dose. This requires a plant to make projected doses for the quarter and year on a 31 day basis. It is only necessary and reasonable to make a projection for the next 31 days. A cumulative dose projection is still required for the current calendar quarter and year in accordance with the ODCH. This is consistent with WOG-72. 5.5 14 ITS Section 5.5.7 is being revised c onsistent withk'0G-35 [ ]. proposed changes to Section 5.5.7 provide an exception to the The ' A 5.5 'L\ examination requirements in Regulatory Guide 1.14. Revision 1. " Reactor Coolant Pump Flywheel Integrity." The proposed exception to the recomendations of Regulatory Position C.4.b would allow for an acceptable inspection method of either an ultrasonic volumetric or surface exarrination. The acceptable inspection method would be enducted at approximately 10 year intervals. This change is consistent with the NRC Safety Evaluation Report associated with WCAP 14i35. " Topical Report on Reactor Coolant Pump Flywheel Inspection Elimination." JUSTIFICATION FOR DIFFERENCES - TS 3 5/15/97

i f CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431. SECTION 5.0 Page 3 of 4 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON C0HANCHE PEAK WOLF CREEK CALLAWAY 5.5-9 Revises Section 5.5.12c to clarify for Diablo Canyon that Yes No No No temporary outdoor liquid radwaste tanks are covered under the surveillance program. This is consistent with current plant practices 5.5 10 This change lists the tanks that the surveillance program No - DCPP CLB does No - Not part of Yes Yes in Section 5.5.12c is applicable to as is in the CTS. This not list tanks. CTS. maintaining change is a plant specific requirement consistent with the ITS wording. CTS. 5.5-11 The documents referenced for the testing frequency for the No - See CN 5.5-12 Yes Yes Yes Ventilation Filter Testing Program (VFTP) do not provide frequencies for combined pressure drop tests or the heater power rating tests. The current IS frequency is added for these tests 5.5-12 The referenced frequencies fur the tests listed in the Yes No No No Ventilation Filter Testing Program (VFTP) were evaluated as part of the 24 month fuel cycle program for DCPP (see LAR 96-09). 5.5-13 Revises Radioactive Effluent Controls Program dose Yes Yes Yes Yes projections to meet original intent of TS prior to

                                                                                                                                                                                                                                                                                      ~1 implementation of GL 89-01. (WOG-72)                                                                                      [73rp_73D                                                       h 5.5-14                    Section 5.5.7 is being revised consistent                                                                     'th    [ ). Yes                                                            Yes                 Yes                            Yes The proposed changes to Section 5.5.7 provide                                                                        tion to the examination requirements in Regulatory Guide 1.14 Revision 1. Reactor Coolant Punp Flywheel Integrity."

5.5 15 This change provides a time interval of within 31 days No No No Yes following removal in which a laboratory test of a sample obtained from the charcoal adsorber must be tested. This change is consistent with the Callaway CTS. 5.6-1 Revises Section 5.6.4. Nonthly Operating Report. for No Yes No No Comanche Peak to reflect a revised submittal date. This change is consistent with the CTS.

5. 5 - 4,

                                                                   .Inse.< t 66-49                                                                                    [cq3.z_({

ce_,em co oamemam , _m.mu CrwuO Lca-sm-04 ,_,

l l i l \ ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 5.5-3 APPLICABILITY: DC, CP, WC, CA l REQUEST: ITS 5.5.4 b&g and Difference 5.5-1 Comment: Changes are based on a yet unnumbered traveler. Use current ITS. FLOG RESPONSE: Traveler TSTF-258 has been submitted to the NRC for review. This traveler superseded travelers, TSTF-86, TSTF-121, and TSTF-167. TSTF-258 is based on the recommendations in the April 9,1997 letter from C. Grimes (NRC) to J. Davis (NEI), with some exceptions. The FLOG submittals have been revised to incorporate TSTF-258. The latest industry status on TSTF-258 is that the NRC has requested changes to l Section 5.7, High Radiation Area. See response to Comment Number , l 5.7-1 for how the FLOG has addressed the NRC comments on TSTF- l 258. I i ATTACHED PAGES: See markups associated with Comment Number O 5.2-1. l l

l l ADDITIONAL INFORMATION COVER SHEET i I ADDITIONAL INFORMATION NO: O 5.5-4 APPLICABILITY: DC, CP, WC, CA REQUEST: ITS 5.5.4 e and Difference 5.5-13 Comment: WOG-72 has not yet t scome a TSTF. Use current ITS. FLOG RESPONSE: This change to ITS 5.5.4 e was prepared in accordance with WOG-72, Rev.1 which was submitted to the NRC and then withdrawn on 2/5/98 by the TSTF to determine if revisions were required. The change specifies that the requirement to determine cumulative dose contributions from radioactive effluents need be done on a current quarterly and annual basis instead of every 31 days. We believe there is a strong technical basis for this change to the ITS and expect NRC approval once it is re-submitted for review. We request that the NRC keep this as an open item under the assumption that it will be approved prior to issuance of the SER. ATTACHED PAGES: None. 1 l l l

! ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 5.5-5 APPLICABILITY: CA REQUEST: ITS 5.5.12 c, CTS 6.8.5 a.3 and Difference 5.5-7 (Callaway) Comment: The CTS just refers to 10 CFR Part 20 Appendix B. More information is needed to determine which table governs the current requirements. , FLOG RESPONSE: FSAR Section 16.11.1.5 provides the controls for the quantity of radioactivity contained in unprotected outdoor liquid storage tanks. The Bases for this section states

" Restricting the quantity of radioactive material contained in the specified tanks provides I assurance that in the event of an uncontrolled release of the tank's contents, the resulting concentrations wou!d be less than the limits of 10 CFR 20.1-20.602, Appendix B, Table 11, Column 2 (re-designated at 56FR23391, May 21,1991) at the nearest potable water supply and the nearest surface water supply in the UNRESTRICTED AREA".

ATTACHED PAGES: None i l l l

ADDITIONAL INFORMATION COVER SHEET l ADDITIONAL INFORMATION NO: O 5.5-6 APPLICABILITY: CA REQUEST: CTS 3.7.6 and Changes 10-15-LG and 1017-A (Callaway) Comment: Please provide a better explanation of the deletion of Pressurization System 2200 CFM +800,-200. FLOG RESPONSE: Surveillance Requirements 4.7.6 c1 and 4.7.6 c3, to verify the system flow rate of 2200 cfm +800, -200 cfm for the Pressurization System has not been deleted it has been moved to the FSAR Section 16.7.8.1.1. DOC 10-15-LG applies to both CTS 4.7.6 c1 and 4.7,6 c3 ATTACHED PAGES: (From Attachment 13) Encl. 2 3/4 7-15

wwLerver 1 e ws PLANT SYSTEMS

SURVE7llANCE RE0UTRFMENTS (Continued)

c. At least once per 18 months, or (1) after any structural maintenance 4 on the HEPA filter or charcoal adsorber housings, or (2) following t b-o 8-A painting, fire or chemical release in any ventilation Zone communicating with the system by:

1) Verifying that the Control Room Emergency Ventilation System so oB-A satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 1% and uses the test procedure guidance in Regulatory Positions C.S.a. C.5.c, and C.5.d of l Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 2000 cfm for the Filtration System and I 500

5-
i 2200 cts, @ Ter the T.asesriz:ti= Syst = .itt. 500 cfm 50 going through the Pressurization System filter adsorber unit;

2) Verifying within 31 days after removal, that a laboratory io-o8-A I analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revis. ion 2, March 1978; meets the laboratore testing ,

criteria of ASTP( D-3803-1989 when tested at 30*t. and 70% relative humidity for a methyl iodide penetration of less than 'm 2%; and Verifying a system flow rate of 2000 cfm for the !

3) to-n-A Filtration System and 2200 T. ' hj Ter tl 2 Tre23srizeticr. go.ts _LG Sy;tes,with cfm going through the Pressurization } 5.5-G 1 System filter adsorber unit during system operation when tested in accordance with ANSI N510-1975.

d. After every 720 hours of charcoal adsorber operation by vehifying s o-oe-A within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of ASTN D-3803-1989 when t Mted at 30'C and 70% relative humidity for a methyl iodide penetration of less than 2%;

e. At least once per 18 months by:

Verifying that the pressure drop across the combined HEPA 10- 09 A 1) filters and charcoal adsorber banks is less than 5.4 inches Water Gauge while operating the system at a flow rate of j . 2000cfm[ fortheFiltrationSystemand500cfm[Soo g for the Pressurization System filter edsorber unit; CALLAWAY - UNIT 1 3/4 7-15 Amendment No. 4 ,106

l l i ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 5.5-8 APPLICABILITY: DC, CP, WC, CA REQUEST: CTS 3.7.6 (3.7.5.1 and 3.7.6.1 - DCPP and 3.7.7.1 and 3.7.8 -CPSES) and Change 10-08-A Comment: It should be specifically noted as to which CTS requirements were carried over to the VFTP and which were deleted (as well as which section of what standard justified the duplication deletions). Provide explanation and justification. FLOG RESPONSE: Attached Table 5.5-8 describes where the CTS SRs for plant ventilation systems were moved to in the ITS. The following provides justification and clarification for those CTS SRs that were not moved to either the "Veatilation Filter Testing Program (VFTP)" in the ITS, or the ITS SRs: . DOC 10-07-LG (not applicable to CPSES) moves the requirement to verify Control Room temperature once every 12 hours to a licensee controlled document. This DOC has been revised to include the fo!!owing additional justification: "The NRC has previously approved moving this type of detailed information or specific requirements to a licensee controlled document that is maintained in accordance with applicable regulatory requirements. This temperature is not an initial condition or controlled ; parameter for any licensing-based accident scenarios Also, its inclusion in the ITS is i not necessary to adequately protect the health and safety of the public. The basic requirements for maintaining OPERABILITY are still retained in the technical specifications." . Per DOC 10-17-A, the SR to measure ventilation system flow rate is not identified as a separate SR in the ITS because it is verified as part of the other in-place filter tests that are specified in ITS 5.5.11. The same DOC applies to CTS SR 4.7.6.1 b 3 for l Diablo Canyon, CTS SR 4.9.13 b 3 for Wolf Creek and CTS SR 4.7.7 b 3 for Callaway for the same reason. . DOC 10-08 A has been revised to show that some CTS SRs were moved to the ITS SRs. ATTACHED PAGES: Table 05.5-8 ITS 3.7 Encl. 3A 12 Encl.38 11

l l l \ l QS.5-8 Response Input 1 TABLki Q5.5-8 I DCPP , WC CA CP . Licensee ) CTS SR ' CTS SR CTS SR CTS SR Controlled ) VFTP ITS SR Document , 4.7.5.1 a 4.7.6 a 4.7.6 a N/A X 4.7.5.1 b 1 4.7.6 b 4.7.6 b 4.7.7.1 a 3.7.10.1 4.7.5.1 b 2 N/A N/A N/A 3.7.10 Bases 4.7.5.1 b 3 N/A N/A N/A 3.7.10 Bases ; 4.7.7.1 b ITS 5.5.11 3.7.10.2 ! 4.7.5.1 c 1 4.7.6 c 1 4.7.6 c 1 4.7.7.1 b 1 ITS 6.5.11a&b l 4.7.5.1 c 2 4.7.6 c 2 4.7.6 c 2 4.7.7.1 b 2 ITS 5.5.11c 4.7.5.1 c 3 4.7.6 c 3 4.7.6 c 3 4.7.7.1 b 3 See DOC 10-17-A l 4.7.5.1 d 4.7.6 d 4.7.6 d 4.7.7.1 c ITS 5.5.11 & 3.7.10.2 l l 5.5.11c l 4.7.5.1 e 1 4.7.6 e 1 4.7.6 e 1 4.7.7.1 d 1 ITS 5.5.11d 3.7.10.2 l 4.7.5.1 e 2 4.7.6 e 2 4.7.6 e 2 4.7.7.1 i 3.7.10.3

i 4.7.5.1 e 3 4.7.6 e 3 4.7.6 e 3 4.7.7.1 j 3.7.10.4 4.7.5.1 e 4 4.7.6 e 4 4.7.6 e 4 4.7.7.1 d 2 ITS 5.5.11e 3.7.10.2 4.7.5.1 f 4.7.6 f 4.7.6 f 4 7.7.1 e ITS 5.5.11 & 3.7.10.2

5.5.11a 1

4.7.5.1 g 4.7.6 g 4.7.6g 4.7.7.1f ITS 5.5.11 & 3.7.10.2 5.5.11 b 4.7.7.1 g ITS 5.5.11 & 3.7.10.2 5.5.11a 4.7.7.1 h ITS 5.5.11 & 3.7.10.2 5.5.11 b 4.7.6.1 a 1 4.9.13 a 4.7.7 a 4.7.8a 3.7.12.1 DC&CP 3.7.12.1 Bases l 3.7.13.1 WC&CA 4.7.6.1 a 2 N/A N/A N/A 3.7.12.1 Bases 4.7.8b ITS 5.5.11 3.7.12.2 4.7.6.1 b 1 4.9.13 b 1 4.7.7 b 1 4.7.8 b 1 ITS 5.5.11a&b 3.7.12.2 DC 3.7.13.2 WC&CA NA-CP I 4.7.6.1 b 2 4.9.13 b 2 4.7.7 b 2 4.7.8 b 2 ITS 5.5.11c 3.7.12.2 DC 3.7.13.2 WC&CA NA-CP 4.7.6.1 b 3 4.9.13 b 3 4.7.7 b 3 N/A See DOC 10-17-A 4.7.6.1 c 4.9.13 c 4.7.7 c 4.7.8 c ITS 5.5.11 & 3.7.12.2 DC&CP 5.5.11c 3.7.13.2 WC&CA 4.7.6.1 d 1 4.9.13 d 1 4.7.7 d 1 4.7.8 d 1 ITS 5.5.11d 3.7.12.2 DC&CP 3.7.13.2 WC&CA 4.7.6.1 d 2 4.7.7 b 2 4.7.7 d 3 4.7.8 d 2 3.7.12.3 vC&CP 3.7.13.3 WC&CA 4.7.6.1 d 3 4.9.13 d 2 4.7.7 d 4 4.7.8 d 3 ITS 5.5.11e 3.712.2 DC&CP 3.7.13.2 WC&CA l_ 4.7.6.1 d 4 N/A N/A N/A 3.7.12.6 3.7.12.6 Bases ! 4.7.6.1 e 4.9.13 e 4.7.7 e 4.7.f. e ITS 5.5.11 & 3.7.12.2 DC&CP ! 5.5.11a 3.7.13.2 WC&CA l 4.7.6.1f 4.9.13 f 4.7.7 f 4.7.8 f ITS 5.5.11 & 3.7.12.2 DC&CP l 5.5.11b 3.7.13.2 WC&CA N/A 4.7.7 b 1 4.7.7 d 2 4.7.8 d 4 3.7.13.4 WC&CA 3.7.12.4 CP

CHANGE NUMBER H5tjC DESCRIPTION inoperable. This change is in accordance with NUREG 1431 and provides clarification only. 10 01 - Not applicable to Callaway. See Conversion Comparison Table (Enclosure 38). i I 10 02 H The APPLICABILITY and applicable ACTIONS are revised to j incorporate "during movement of irradiated fuel ; assemblies" in addition to all MODES (i.e.. MODES 16). 10 03 - Not applicable to Callaway. See Conversion Comparison Table (Enclosure 38). 10 04 A A new ACTION statement is added by NUREG 1431 to require ! entering TS 3.0.3 immediately if two trains of the CR ventilation system are inoperable in MODES 1, 2. 3. or 4. The CTS requires entry into TS 3.0.3, since the condition of two trains inoperable is undefined; therefore, the revision has been classified as administrative. 10 05 LS 18 A ncw option is added to the ACTION by NUREG 1431 that allows the suspension of CORE ALTERATIONS or movement of irradiated fuel versus placing the CR ventilation system i in the recirculation mode. 10 06 LG The details and description of the required ACTIONS and the monthly SRs for train operability are moved to the l Bases. This is an example of removing details that are i not required to be in TS and is consistent with i NUREG 1431. Rev. 1.

                                                                                                                                 .g n w Q                          ;Q 5.5-8) 10 07                                       LG E           ur                 if s                       r C

10 08 A The description of the ventilation filter specific testing l l l requirements and the required surveillances are moved to the Ventilation Filter Testing Program (VFTP) as defined in the Adninistrative Controls of the ITS. No technical changes to requirements or test specifics except as noted in separate change ntabers are made. A new SR is added that requires [ Control Room Emergency Ventilation and Emergency Exhaust] system filter testing in accordance j Q 5.5-B

( with the VFTP. The requirements of this specification 1 . are: 1)movedtoSection5.5.11oftheITS.or2)picted i- :ir.:: t M y ve 9171icated in ^ W etery Cuide 010)31.52 l Jt:visier. 2. [ ANSI N510-1975, er ASTM D ^003 1^^^]. f ( ~ hvdh_rTSSRs.)

DESCRIPTION OF CHANGES TO CURRENT TS l'2 5/15/97 1_- -. - - . _. . _ . . - .

Insert for Q5.5-8 Enclosure 3A of CTS 3/4 -7, Page 12 INSERT for DOC 10-07-LG i The surveillance that verifies CR temperature once per 12 hours is moved to a licensee controlled document. The NRC has previously approved moving this type of detailed information or specific requirements to a licensee controlled document that is maintained in accordance with applicable regulatory requirements. This temperature is not an initial condition or controlled parameter for any licensing-based accident scenarios. Also, its inclusion in the ITS is not necessary to adequately protect the health and safety of the public. The basic requirements for maintaining OPERABILITY are still retained in the technical specifications. l 1 I l I l i i 1 l 1 i )

l CONVERSION COMPARISON TABLE - CURRENT TS 3/4.7 Page 11 of 15 TECH SPEC CHANGE APPLICABILITY DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY NUMBER DESCRIPTION The surveillance that verifies CR temperature once per YES: move to ECGS. NO: not in CTS. YES: moved to USAR. YES: moved to FSAR. 10-07 LG 12 hours is noved to a licensee-controlled document. r YES YES 10 08 The description of the ventilation filter specific testing YES YES A requirements are moved to the VFTP. as defined in the Ackninistrative Controls of the ITS.W or it:ted :: ht ; ~ M h IT55R]s --@ 5.5-9 I

          &p' i + ~4 i a +" :;;1'ed?: 9 y StM-6. A SR is added that requires [ Control Room Emergency Ventilation and Emergency Exhaust System] filter testing in accordance with the VITP.                                               -

NO. Refer to NO: not in CTS. YES YES 10-09 The action for an OPERABLE ventilation train not being I capable of being supplied from an emergency power source is change 10-16 LG. LS-27 deleted. YES YES YES YES 10-10 The SR is revised to allow credit for an actual actuation TR-1 and moves signal specifics to the Bases. YES YES YES 10-11 Frequency of the surveillance requiring verification of the YES LS-19 control room ventilation system capability to maintain a positive pressure in the CR is relaxed to 18 months on a STB. NO: not in CTS. YES YES 10-12 De!.a= the STB for the 31 day testing. NO: CTS LS-32 surveillance is not STB. NO NO NO 10-13 The DCPP specific footnotes indicating the control roon YES LG ventilation system is consnon to both units and that the system may be considered OPERABLE with no chlorine monitors if no bulk chlorine gas is stored within the SITE BOUM)RY. are moved to the Bases. NO: TS 3.0.4 YES YES 10-14 The statement that LCO 3.0.4 is not applicable is deleted NO: TS 3.0.4 A based upon the new ITS definition of LCO 3.0.4 which does exemption is not in exesption is not in not apply in MODES 5 and 6. CTS. CTS. 5/I5/97 CONVERSION COMPARIS0N TABLE - CURRENT TS

l 1 l l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 5.5-9 APPLICABILITY: DC, WC, CA REQUEST: CTS 3.9.13 (3.9.12 - DCPP) and Change 12-04-A (Wolf Creek, Callaway and Diablo Canyon) Comment: It appears that some of the CTS requirements covered by this change were deleted rather than transferred to ITS 5.5.11 as stated. Justify the individual deletions. FLOG RESPONSE: Attached Table 5.5-9 describes where the CTS SRs for fuel building ventilation systems were moved to in the ITS. The following provides justification and clarification for those CTS SRs that were not moved to either the Ventilation Filter Testing Program (VFTP) in the ITS, or the ITS SRs:

Per DOC 12-11-A, the SR to measure FHBVS flow rate is not identified as a separate SR in the ITS because it is verified during the other in-place filter tests specified in ITS 5.5.11, " Ventilation Filter Testing Program (VFTP)", and specific ITS SRs. This change does not result in a change to the technical requirements.

DOC 12-04 has been revised to more clearly describe where the CTS SRs were moved to in the ITS. l ATTACHED PAGES: Table Q5.5-9 ITS 3.9 Encl. 3A 8 l 1 l t l

l 1 l l l QS.5-9 Response input l TABLE Q5.5 9 l DCPP WC CA CP-l CTS SR CTS SR CTS SR CTS SR VFTP ITS SR I 4.9.12 a 4.9.13 a 4.9.13 a N/A 3.7.13.1 4.9.12 b 1 N/A N/A N/A 3.7.13.5 4.9.12 b 2 4.9.13 b 1 4.9.13 b 1 N/A ITS 5.5.11a &b 3.7.13.2 l 4.9.12 h .3 4.9.13 b 2 4.9.13 b 2 N/A ITS 5.5.11c 3.7.13.2 l 4.9.12 b 4 4.9.13 b 3 4.9.13 b 3 N/A See DOC 12-11-A ! 4.9.12 c 4.9.13 c 4.9.13 c N/A ITS 5.5.11c 3.7.13.2 4.9.12 d 1 4.9.13 d 1 4.9.13 d 1 N/A ITS 5.5.11d 3.7.13.2 4.9.12 d 2 4.9.13 g 1 4.9.13 d 2 N/A 3.7.13.3 4.9.12 d 3 4.9.13 g 2 4.9.13 d 3 N/A 3.7.13.4 DC 3.7.13.5 WC&CA 4.9.12 e 4.9.13 e 4.9.13 e N/A ITS 5.5.11a 3.7.13.2 l 4.9.12 f 4.9.13 f 4.9.13 f N/A ITS 5.5.11b 3.7.13.2 N/A 4.9.13 d 2 4.9.13 d 4 N/A ITS 5.5.11e 3.7.13.2 WC&CA l l l l 1 l l t i 4 i

l CHANGE NUMBER tGiG DESCRIPTION 11 04 LG This change moves the restriction on crane operation to a licensee controlled document. The restriction on crane operations may be removed because it is not in the assumptions used for the FHA. Crane operations that could adversely affect fuel stored in the spent fuel pool are , controlled as analyzed in the review of heavy load I movements. This change is consistent with NUREG 1431. J Rev. 1. and moves requirements that do not meet the i criteria for inclusion in the TS. 12 01 LS 24 The applicability would be changed to "During movement of l irradiated fuel in the fuel building" instead of "Whenever i irradiated fuel is in the spent fuel pool." consistent i with NUREG 1431. Rev. 1. The proposed applicability is consistent with the asstaptions used in the Fuel Handling Accident in the fuel building which postulates the inadvertent drop of an irradiated fuel assembly. Potential damage to fuel assemblies due to dropping of heavy loads is addressed by change 12 02 LG. 12 02 LG Hoves the restriction on crane operations over the spent fuel storage areas when the fuel building air cleanup system was inoperable. The restriction on crane operations may be moved because it is not consistent with the assumptions used for the FHA. Crane operations that could adversely affect fuel stored in the spent fuel pool are prohibited in accordance with plait procedures as analyzed in the review of heavy load movements. 12 03 A The statement that 3.0.3 [and 3.0.4] are not applicable would be removed. This is consistent with the proposed change to integrate the emergency exhaust system requirements for irradiated fuel handling in the fuel l building with the emergency exhaust system requirements in Modes 1 through 4. ITS 3.7.13 supports this integration of requirements. - 12 04 A The S rv llan e R ir nts egard'ng f ite tes ing Q 5 5-% f woul mov to "Ven ilat n F1 ter est g P ogr m" th i call ou in a ini rati e co tro ti 5 .1 of t IT . Th's ch ge d s no re uit in. an to chni al r uir nts.

InSe r 6 12 05 TR 1 Revised Surveillance Requirement to allow for increased flexibility in using an actual or simulated actuation signal . Identification of the specific actuation signal is moved 20 the Bases.

DESCRIPTION OF CHANGES TO CURRENT TS 8 5/15/97

i l Insert for QS.5-9 A of CTS 3/4 - 9, Page 8 INSERT for DOC 12 04: Chance Number Description 12-04 The description of the ventilation filter specific testing requirements and the A required surveillances are moved to the " Ventilation Filter Testing Program (VFTP)* as defined in ITS 5.5.11. No technical changes to requirements or test specifics except as noted in separate DOCS are made. A new SR is added that requires [ Emergency Exhaust System] filter testing in accordance with the VFTP. The requirements of this specification are either : 1) moved to ITS 5.5.11, or 2) moved to ITS SRs.

l ADDITIONAL INFORMATION COVER SHEET i ADDITIONAL INFORMATION NO: O 5.5-10 APPLICABILITY: WC, CA REQUEST: ITS 5.5.11.b (Callaway and Wolf Creek) Comment: The smooth copy of the ITS still has the [] around the plant specific bypass value FLOG RESPONSE: The smooth copy of the ITS has been mar'<ed to delete the brackets ([ ]) around the plant specific bypass '/alue. A final review of the smooth ITS and ITS Bases is planned prior to resubmitting to the NRC the smooth copy of the ITS and Bases ATTACHED PAGES: None

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 5.6-1 APPLICABILITY: DC, CP, WC, CA REQUEST: ITS 5.6.5 a.7&8, Changes 03-14 & 15 M Comment: It is true that the additions would make the COLR more restrictive however, the removal of the specific values from the TS is a less restrictive change that needs to be justified. Provide justification. FLOG RESPONSE: DOC-03-14-M describes the addition of the SHUTDOWN MARGIN (SDM) limits and the Moderator Temperature Coefficient (MTC) limits to the Administrative Program description of the CORE OPERATING LIMITS REPORT (COLR). As stated, this change is more restrictive to the COLR. The less restrictive change of moving the actuallimits from the technical specifications to the licensee controlled COLR are addressed and justified by DOC 01-01-LG (SDM) found in Section 3.1 (not applicable to CPSES) and DOC 03-07-LG (MTC) found in Section 3.1 (applicable to DCPP only). DOC-03-15-M, in a similar way, adds the Refueling Boron Concentration limits to the Administrative Program description of the COLR. The less restrictive change of moving these limits to the licensee controlled COLR is addressed and justified by DOC 01 LG found in Section 3.9. ATTACHED PAGES: None i

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

l l . l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: O 5.7-1 APPLICABILITY: DC, CP, WC, CA REQUEST: ITS 5.7.2 and Difference 5.7-2 l Comment TSTF-167 has been rejected by the NRC. Use current ITS. FLOG RESPONSE: Traveler TSTF-258 has been submitted to the NRC for review. l l This traveler superseded travelers, TSTF-86, TSTF-121, and TSTF-167. TSTF-258 is based on the recommendations in the ) April 9,1997 letter from C. Grimes (NRC) to J. Davis (NEI), with I some exceptions. The FLOG submittals have been revised to incorporate TSTF-258 and encompass the NRC comments of ! 6/11/98. Additional technical changes made to Section 5.7 are l identified and justified. (See JFD 5.7-2 which revises ITS 5.7.2e l consistent with CTS 6.12 and JFD 5.7-4 which revises ITS 5.7.2f consistent with CTS 6.12.) The latest industry status on TSTF-258 is that the NRC has requested changes to Section 5.7, High l Radiation Area. ATTACHED PAGES: See markups associated with Comment Number O 5.2-1. l I 1 j 1 i I

! ADDITIONAL INFORMATION COVER SHEET ! ADDITIONAL INFORMATION NO: CA 5.0-002 APPLICABILITY: CA REQUEST: Reflect changes made by OL Amendment No.122, dated 3/23/98. 1 ATTACHED PAGES: Encl. 2 6-1,6-14,6-15 l l l l 1 I l i

                                                     ,-w,-

l l l j ADMINISTRATIVE CONTROLS l 6.1 RESPONSIBILITY j l .6.1.1 The Manager. Callaway Plant, shall be responsible for overall unit l operation and shall delegate in writing the successicn to this responsibility i during his absence. 1 (iM<tL}6.1.2The Shift Supervisor (cr dur"; hi: 2:ent: from a,: control co;" 6 L de:1:ncted individual) shall be responsible for the control rcam command ol-ol- A Nd Tunction3-A m:n ;ryt directive to thi Offect. ;gned b., the V ue Pcea, e

:nd Chief Nuck;; Or'icer shall te reissued to ' station pcr:cnncl en--m Ennuel basis.

J 6.2 ORGANIZATION 6.2.1 Onsite and Offsite Oraanization An casite and offsite organization shall be established for unit operation and corporate management. The onsite and offsite organization shall include the positions for activities affecting the s~fety a of the. nuclear power plant.

a. Lines of authority, responsibility and communication shall be established and defined for the highest management levels through
intermediate levels to and including all operating organization t

positions. These relationships shall be documented and updated. I I as appropriate, in the form of organizational charts, functional descriptions of departmental responsibilities and relationships. and job descriptions for key personnel positions or equivalent forms of documentation. These' requirements shall be documented in the FSAR. nd updated " ::: rtn:: with 10 C.'R 30.71 4 . o(-of- A

b. The Manager. Callaway Plant shall have responsibility for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant. ,9, ,

c. The Vice President and Chief Nuclear Office; shall have corporate l g responsibility for overall plant nuclear safety and snall take any l measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support in the plant to ensure nuclear safety.

d. The individuals who train the operating staff and those who carry out the health physics and quality assurance functions may report to the :-ppropriate onsite manager: however, they shall have

! sufficient organizational freedom to ensure their independence j from operating pressures. l s i j CALLAWAY - UNIT 1 61 Amendment No. 2'.29,122 i _. _. - . . _ __ ._ _ ~ . _ .-

CA-T.0 - 001 \ INSERT 1 ' i The' Manager, Callaway Plant or his designee shall approve, prior to implementation, each proposed test, experiment or l' modification to systems or equipment.that affect nuclear OlsDl- A safety and are not addressed'in the Final Safety Analysis Report (FSAR) or Technical Specifications. ,

INSERT 2 During any absence of the SS from the control room while the unit is in MODE 1, 2, 3, or 4, an individual with an active Senior Reactor Operator '(SRO) license shall be designated to assume the control' room command function. During any Ol-o l- A absence of the SS-from the control room while the unit is in j MODE 5 or o, an individual with an active SRO license or J Reactor Operator license shall be designated to assume the control room command function. 1

,e e l 1 __-i

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ADMINISTRATIVE CONTROLS C A- L o - o o *2- \

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l. 68 PROCEDURES AND PROGRAMS l

6.8 1 Written procedures shall be established. implemented, and maintained i coverint the activities referenced below: i s

a. The applicable procedures recommended in Appendix A. of Regulatory l f Guide 1.33. Revision 2. February 1978;

, b. The emergency operating procedures required to implement the requirements.of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Section 7.1 of Generic Letter t.o. 82-33: n,,..... ., .e _ _, ,- ... 4 . ..s . o,m -7. . -

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g. Quality Assurance Progra. ' amentation for effluent and environmental monitorin .-aM.

                                  -h.                     Fire Protection Program implementati > n.n b c e 6        7,-L                  ~ - _ A.            --
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6.8.4 The following programs shall be established. implemented, and maintained: ( QI<t ( page (.-2.5 , pnpph G.14. h<rt.

a. Reactor Coolant bources Outside Containment ol-oi-A i l

A program to reduce leakage from those portions of systems outside l containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. T' 9 l- i l' l f systems include-the recirculation portion of the Containment Sprc; System. Safety Injection System. Chemical and Nolume Control System. and RHR System. The program shall include the following: r-L 1) Preventive maintenance and periodic visual inspection i

3 requirements, and 4 m - l l i Au prop.ms spdAu 6 Soc 1c6 CA.4 c4 G .t.5. 02.-o9 M CALLAWAY - UNIT 1 6-15 Amencment No. 20.22.102.; ) 122 v.-p,. m . , , . r. , - - - , - -

                                                                                                                        .mr                         ,.,c                                                                                         *-a

l l l ADDITIONAL INFORMATION COVER SHEET i ADDITIONAL INFORMATION NO: CA 5.0-003 APPL!CABILITY: CA, DC, WC i REQUEST: ITS 5.5.10, is revised to delete the words: "and low pressure turbine disc stress corrosion cracking" This requirement is not part of the Secondary Water Chemistry Program described in CTS 6.8.4c. ATTACHED PAGES: Encl.5A 5.0-23 Encl. 6A 4 Encl. 6B - 3 i l i

I Programs and Manuals 5.5 i 5.5 Programs and Manuals (continued) 5.5.10 Secondary Water Chemistry Proaram (C A - 5. 0 -oo 3 This program provides controls for monitoring secondary water chemistry _ to i it tube degradationfTnd Tir.: pre;;0rc turbine di;c ;trc;; corrc;ica e program shall include: 5- W

a. Identification of a sampling schedule for the critical variables and control points for these variables:

b. Identification of the procedures used te measure the values of the critical variables:

c. Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in leakage;

d. Procedures for the recording and management of data;

e. Procedures defining corrective actions for all off control point chemistry conditions; and

f. A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action.

5.5.11 Ventilation Filter Testina Procram (VFTP) A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems B PS at the frequencies specifiet in Regulatory Guide 1.52CRevi7. and in accord;ncc with Uses"thettestrprocedure; guidance in ERegulatory Guide 1.52, Revision 2 Positions?C 51aQC.'5;c;and'C05;d PS A li: l1:10 1000, and AC 1].

a. Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a B .PS penetration and system bypass < 1:0* when tested in accordence with [Rc;iulatory Cuid 1.02. R;vi; ion 2. end ADiC l:101000] at PS the system flowrate specified below [110t].

l l (continued) MARX UP OF WOG STS REV 1 (NUREG 1431) 5.0 23 5/15/97 l i

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.10 Secondary Water Chemistrv Proaram 5. o - oo 3 ((, A This program provides controls for monitoring secondary water chemistry _ to inhibit 3 tube degradationj$d 13 ;irc:: Orc _ turbine disc strcss crrc; ion he program shall include: 5- U

a. Identification of a sampling schedule for the critical variables and control points for these variables:

b. Identification of the procedures used to measure the values of the critical variables:

c. Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in leakage;

d. Procedures for the recording and management of data:

e. Procedures defining corrective actions for all off control point chemistry conditions; and

f. A procedure identifying the authority responsible ~for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action.

5.5.11 Ventilation Filter Testir,a Proaram (VFTP) A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems B.PS at the frequencies specified in Regulatory Guide 1:52ERev.~2. and +n accordance with uses4he'testiprocedure.; guidance in ERegulatory Guide 1.52. Revision 2. Positions'C:5'.a',1CJ5:cpand'C25~.d PS A2i: N510-1000, and AC 13

a. Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a _ B PS penetration and system bypass < 1.~0% when tested in accordancc with [ Regulatory Guidc 1.52 ncvision 2. ;>nd A21: N510-1000] at PS the system flowrate specified below [110i].

(continued) MARK UP 0F WOG STS REV 1 (NUREG 1431) 5.0 23 5/15/97

CHANGE Mit1Bf8 JUSTIFICATION 5.5 15 ITS Section 5.5.11c is being revised consistent with the CTS. The proposed changes specifies a time interval "of within 31 days after removal" in which a laboratory test of a sample obtained from each of the ESF systems charcoal adsorber must be tested for methyl iodide

                                   ,1.               A- 4kh                @        'I 5.6 1         Not applicable to Callaway Plant. See Conversion Comparison Table (Enclosure 6B).

yg I CA -5.o- oo3 5.6 2 This change deletes the EDG Report to reflect the recommendations of , GL 94 01, " Removal of Accelerated Testing and Special Reporting Requirements for Emergency Diesel Generators," dated May 31, 1994. 5.6 3 This change revises the report date in Section 5.6.P. ' Annual Radiological Environmental Operating Report" to be consistent with the CTS. 5.6 4 This change revises Sections 5.6.1 and 5.6.3, " Occupational Radiation Exposure Report" and " Radioactive Effluent Release Report," respectively, per NRC letter dated July 28, 1995, " Changes to Technical Specifications Resulting from 10 CFR 20 and 50.36a Changes" (From

Christopher I, Grimes to Owners Groups Chairs). This is consistent with traveler TSTF 152. 5.6 5 dd=PORV o- lift settings are referenced in PTLR section perQ TR 5.o-oo3)

h. G -(, In u r t (,A 4b ) MQ 5.2-1 [x 5.7 1 (This ange vises igh iaJt'e6Are/ao ipedrpojr t'e Jc Wigefs co iste with 0 CF 0.16U1). Sp6cifically,41istaffces Momfttfe adia n sour e. :En A f t G A - 4 C- ___ ,

Q 5.2- 1 ~

5.7 2 Thi ha g rev' es "u uthor' ed" t "ina rt "i the igh diat'6n Ar sect' n to flec eN s siti c; tat in/ RG 8. S tion .5 re rain hysi bar iers or gh diaMon Qras,_ is i cons' tent { h tr eler STF- . Insert 6 A 5.7-3 ITS Section 5.7.3 is being revised consistent with the CTS. The _ proposed change deletes the phrase "or that cannot be continuously guarded" from the ITS to be consistent with the practices of the current TS which do not have~this' requirement for high radiation areas. .--- - - - - - - ~~ '- The use of barricades and flashing lights is adequate protection for individual high radiatior, areas to limit personnel accessibility. 5.F 4 In5u t GA -

                                                                                        $ 5.2-1              3 kmC c-C A-5.o- oc JUSTIFICATION FOR DIFFERENCES          TS            4                                                      5/15/97

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

I CA5o-oo3[ Insert for Encl. 6A Sht 4 ~ 5.5-17 ITS Section 5.5.10 is being revised consistent with CTS 6.8.4.c. The proposed change deletes the phrase "and low pressure turbine disc stress corrosion cracking" from the ITS to be consistent with the practices of the CTS which do not have this requirement for the Secondary Water Chemistry Program. l l 1 l l y

CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431. SECTION 5.0 Page 3 of 4 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY 5.5-9 Revises Section 5.5.12c to clarify for Diablo Canyon that Ye: No No No temporary outdoor liquid radwaste tanks are covered under the surveillance program. This is consistent with current plant practices. 5.5-10 This change lists the tanks that the surveillance program No - DCPP CLB does No - Not part of Yes Yes in Section 5.5.12c is applicable to as is in the CTS. This not list tanks. CTS. maintaining change is a plant specific requirement consistent with the ITS wording. CTS. 5.5-11 The documents referenced for the testing frequency for the No - See CN 5.5-12 Yes Yes Yes Ver.+11ation Filter Testing Program (VFlP) do not provide frequencies for combined pressure drop tests or the heater power rating tests. The current TS frequency is added for these tests. 5.5-12 The referenced frequencies for the tests listed in the Yes No No No Ventilation filter Testing Program (VFlP) were evaluated as part of the 24 month fuel cycle program for DCPP (see LAR 96 09). 5.5-13 Revises Radioactive Effluent Controls Program dose Yes Yes Yes Yes projections to meet original intent of TS prior to implementation of GL 89 01. (WOG-72) [Tsrp _23g O55-1 5.5-14 Section 5.5.7 is being revised consistent h M [ ]. Yes j

                                                                                                                                                         / Yes                 Yes         Yes The proposed changes to Section 5.5.7 provide 3tteaception to tha examination requirements in Regulatory Guide 1.14 Revision 1. Reactor Ccolant Pump Flywheel Integrity."

5.5-15 This change provides a time interval of within 31 days No No No Yes following removal in which a laboratory test of a sarple obtained from the charcoal adsorber must be tested. This change is consistent with the Callaway CTS. 5.6-1 Revises Section 5.6.4. ' Monthly Operating Report. for No Yes No No Comanche Peak to reflect a revised submittal date. This change is consistent with the - 5 . 5 - U, Insert G6-49 [cq s.b l { ce _ ,emco-o,comum . .mu CrmuO _ lca-s.od s , m ,,,

I 2 C A -5.o - 0o 3 Insert for Encl. 68 Sht 3 DC CP WC CA 5.5-17 This change deletes the phrase "and icw YES NO YES YES pressure turbine disc stress corrosion cracking" from ITS 5.5.10 to make the program consistent with CTS 6.8.4.c. j l l l i I. 1 I i U d t i k 1 e we y--e -- -- ,, e m+,v w - --- ----

ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: CA 5.0-004 APPLICABILITY: CA REQUEST: Add to CTS 6.9.1.9 and ITS 5.6.5 the reference to WCAP-12610 as an analytical method used to determine the core operating limits. Amendment No.110 for Callaway was issued on April 30,1996. This amendment modified CTS 5.3.1, " Fuel Assomblies" to add ZlRLO at feel clad material and allow the use of ZlRLO filler rods in piace of fuel rods. The accompanying safety evaluation for the amendment cites topical report WCAP-12610, " VANTAGE + Fuel Assembly Reference Core Report" as the basis for allowing the use of ZlRLO clad fuel rods and ZlRLO filler rods, therefore the addition of this WCAP to the CTS 6.9.1.9 and ITS 5.6.5 is needed. ATTACHED PAGES: Encl. 2 6-21a Encl.3A 10 Encl. 3B 7 Encl. 5A 5.0-34 Encl. 6A 4 Encl.6B 4 l i l i i i f

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

ADMINISTRATIVE CONTROLS

b. WCAP-30216-P-A, REV. l A, " RELAXATION OF CONSTANT AXI AL Of f SET CONTROL AND FQ SURVEILLANCE TECHNICAL SPECIFICATION." February 1992 (W Proprietary) .

l (Methedclegy for Speci#icatier 2.2.1

                                                                                                                                          \
                                                                                                     ^~ial c1= Di'fer n::                 i (Pelaxed Avi:1 0#f et Centrel) and 2.2.2 " :: cl u:: "ct 0 3-l~l- A, Channel F::ter (W(?) :urvei'l:nce requi ement: O f F, o_.u_2_,__.a a as bilvw v s wyj j

c. WCAP-10266-P-A, REV. 2, "THE 1981 VERSION OF WESTINGHOUSE EVALVATION MODEL USING BASH CODE." March 1987 (!! Proprietary).

(Mathed;1cgy for Specific tie- 2.2 2. "c;; ria ";; Char :' i _ . _ _1

                                             .e-. ,.                                                                         o*b-l7- A  !

Jnkib S A ,

                                                                                                                      /C A -5.o -oo4d Tne core operating limits shall be determined so that all applicable limits (e.g. , fuel thermal-mechanical limits. core thermalhydraulic limits, nuclear y

limits such as shutdown margin. and transient and accident analysis limits) cf Ine safety analysis are met. The CORE OPERATING L]MITS REPORT, including any mid-cycle revisions or supplements shall be provided, upon issuance for each reload cycle. to tr.e !.:,', Occu.crt Cort r:1 Dc:' r th copie! tc the o e;ic"

                                                        '      4
                                                                                              ' ^d-iritteter 2nd *: 40:

A =::::r. 0348-A h6Se(6 6 4'I

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03 -oB- A

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ed
r t he Oe;4 n 1 *d-' 4:t :t-: :'

th !!RC " ;;i cn:1 Of # ice M + hi- the t% period :peci' icd for :::b report ! I

         . . 1. r.5        D l' f.* A. D N. D T. T T t11 T. Aff g3-gg.[4 Ir, ;dditicr 10 th Opplic:ble ::Ord retortier requirem:nt: of Ti;le 10, O;e j

of Federal Regul: tion:, the fcil: i L ng rc Ord: :hal' be ret:ined for at least ne nn..i.T.s, period indi::ted. 6.10.1 Th0 f Ol '. OW4 ; rO Ord :hal' b Tel in:d f0r ;l 1 L-t : f ili 3.

a. Recards a~d log; cf unit Oper: tion 0. ring tim: .nter ci at wa:P ps v vv E 4 swwwa,

( CALLAWAY - Ut117 ] 6-21a Amendment No.-C C Z 3 W

INSERT SA , , l

d. NRC Safety Evaluation Reports dated July 1,1991, " Acceptance for Referencing of Topical Report WCAP-12610 ' VANTAGE + Fuel Assembly Reference Core Report' (TAC NO 77268)," and September 15,1994, " Acceptance for Referencing of Topical Report WCAP-12610, Appendix B, Addendum 1,

      ' Extended Buraup Fuel Design Methodology and ZlRLO Fuel Performance                        !

Models' (TAC No, M86416)" (WCAP-12610-P-A) F' l 4 i

CHANGE NUMBER NSHC DESCRfPT10N 3 inadvertent entry is discussed in section 1.5 of RG 8.38.

    -)
 ' *O This RG reflects the NRC's position regarding physical barriers for high radiation areas. Radiation areas within the limits listed shall be locked or continuously guarded to prevent inadvertent entry as discussed in RG 8.38.

Furtheraore, the distinction between unauthorized versus inadvertent is important based on a Notice of ViolatiJn that Callaway received on this interpretation of terms. 03 20 LS 3 Not applicable to Callaway Plant. See Conversion Comparison Table (Enclosure 38). Ifiitvf, \ CA-5. 0-004) J l l i 4 2 DESCRIPTION OF CHANGES TO CURRENT TS 10 5/15/97

D ~ OU Insert for Encl. 3A Sht 10 03-21 A Add to CTS 6.9.1.9 the reference to WCAP-12610 as an analytical method used to determine the core operating limits. Amendment No.110 for Callaway allowed the use of ZlRLO clad fuel rods and ZlRLO filler rods based on WCAP-12610. The addition of this reference to the CTS is app opriate based on the recent issuance of Amendment flo.110 for i WCNOC which added the reference to CTS 6.9.1.9 per the NRC Staff , request. l l l l l 1

CONVERSION COMPARISON TABLE - CURRENT TS 6.0 Page 7 of 7 TECH SPEC ChMGE APPLICABILITY r DIABLO CANYON COMANCHE PEAK WOLF CREEK CALLAWAY NUMBER DESCRIPTION Yes No - Already part Yes Yes 03-14 Shutdown Margin values would be moved to COLR per traveler of CTS i H TSTF-9. In addition. moderator tesperature coefficient limits would also be relocated to the COLR. , Adds refueling boron concentration limits to COLR. Yes Yes No - already in CTS Yes 03-15 H No Yes No No 03-16 Deletes one of the allowed ECCS evaluation models for CPSES A Unit 2 which is no longer used. No - References do Yes Yes Yes 03-17 Deletes the methodology section references in the COLR. A not exist in DCPP CTS La No Yes No 03-18 Hoves the reporting requirement for documentation of all A challenges to the PORVs or safety valves to the Wolf Creek Honthly Operating Report. Yes Yes Yes 03-19 The term " unauthorized ~ is changed to " inadvertent" in the Yes A High Radiation Area section. The prevention of inadvertent entry is discussed in section 1.5 of RG 8.38. Yes Yes No: maintaining CTS 03-20 The use of a continuous guard is provided as an additional Yes LS 3 option for preventing inadvertent entry into high radiation areas that are accessible to individuals. . . . . B

* ~

l Insert for Encl. 3D Sht 7 DC CP WC CA l 03-21 This change adds the reference to WCAP- NO NO NO YES A 12610 to CTS 6.9.1.9 as the basis for allowing !. the use of ZlRLO clad fuel rods and ZlRLO , fi!ler rods per Amendment No.110 for

l. Callaway. ;

1 l 1 l i I l L p l i 4 f 1 5-1

                             -    , . - -       y                 <-.,er m-                                     - - . w     -       -               -----*,,(

I Reporting Requirsments 5.6 5.6 Reporting Requirements 1 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) i

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents: _

Tdcatify th; Topical Report's) by nus.bcr. titic, datc. and NRC B PS staff 7 proval docu . cat, or idcr,tify the staff C;ftty Cvelustica Repsrt for ; plant opccific s,ethodology by NRC ictter end dotc. 1: WCAP 9272 P A. " WESTINGHOUSE ~ RELOAD', SAFETY' EVALUATION l HETH000 LOGY" fJuly 19851([ Proprietary). l

2. WCAP 10216;P A,(REY: 11A',':"RELAXATIONf0F CONSTANT:~ AXIAL OFFSET ~CONTROLo ANDJFQ;SURVEILLAN.CEHECHNICAL'SPECIFICATI0N'"

February E1994T(WProptietary);

3. WCAP710266]EAEREY '?.;7"THE;1981?VER3 ION;0FjWESTINGH0VSE i EVALUATIONJHODELiUEING~;BASHCODE;EHarchy1987 (ce-5.o-co4)

(FProprietary)~. gg, 5.G ~7-

c. The core operating limits shall be determined such that all l applicable limits (e.g., fuel thermal mechanical limits, core ;

thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and I accident analysis limits) of the safety analysis are met,

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Reactor Coolant System (RCS) PRESEBE AND TEMPERATURE LIMITS REPORT B PS (PTLR)

a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, c:A hydrostatic testing and PS PORV lift setting as well as heatup and cooldown rates shall be 5.6 5 established and documented in the PTLR for the following:

1. Specification 3.4.3, "RCS Pressure and Temperature (P/T)

Limits," and 1 (continued) MARK UP OF WOG STS REV 1 (NUREG 1431) 5.0-34 5/15/97

INSERT 1 CA 5.0 - 004-]

4. NRC Safety Evaluation Reports dated July 1,1991, " Acceptance for Referencing of Topical Report WCAP-12610 ' VANTAGE + Fuel Assembly Reference Core Report' (TAC NO 77268)," and September 15,1994, " Acceptance for Referencing of Topical Report WCAP-12610, Appendix B, Addendum 1, .

          ' Extended Burnup Fuel Design Methodo'ogy and ZlRLO Fuel Performance                                      i Models' (TAC No. M86416)" (tNCAP-12610-P-A)                                                               {

l l I s

CHANGE NUMBER JUSTIFICATION 5.5 15 ITS Section 5.5.11c is being revised consistent with the CTS. The proposed changes specifies a time interval "of within 31 days after removal" in which a laboratory test of a sample obtained from each of the ESF systems charcoal adsorber must be tested for methyl iodide DenetratioIL - F N$ 5.5-l(o ~I

                    .En s< r E G A- 4-dh                               -

5.6 1 Not applicable to Callaway Plant. See Conversion Comparison Table (Enclosure 6B). yg{ [ CA-5.o- oo3 l 5.6 2 This change deletes the EDG Report to reflect the recommendations of GL 94 01 " Removal of Accelerated Testing and Special Reporting Requirements for Emergency Diesel Generators." dated May 31, 1994. 5.6 3 This change revises the report date in Section 5.6.2, " Annual Radiological Environmental Operating Report" to be consistent with the CTS. 5.6 4 This change revises Sections 5.6.1 and 5.6.3, " Occupational Radiation Exposure Report" and " Radioactive Effluent Release Report," respectively, per NRC letter dated July 28,1995 " Changes to Technical Specifications Resulting from 10 CFR 20 and 50.36a Changes" (From Christopher I, Grimes to Owners Groups Chairs). This is consistent with traveler TSTF 152. - 5.6 5 PORV lift settings are referenced in PTLR section per@ o_ s TR 5.o-co3) h.G-G In urt GA. 4b) d{ M"b 5.7 1 (This ange vises igh Jadia pt Are/aoipedrport'eJcarfgefs co aiste with 0 CFIVl0.lt#1]. Sp6cifWally, istafices ffom,We l adia n sour e. 1054ft 6A-4C. -- 5.7 2 Thi ha g rev' es "u uthor' ed" t "ina rt "i the igh/ at'on Ar sect' n to flec eN s siti as tat in

                . RG          8, S tion .5 re rdin          hysi      bar iers or            gh    ia on Qr as,             is i cons' tent        h tr eler STF-            . ISSeft 6A-4dj 5.7 3          ITS Section 5.7.3 is being revised consistent with the CTS. The proposed . change deletes the phrase "or that cannot be continuously guarded" frca the ITS to be consistent with the practices of the
            ~~ - current TS which do not have~this requirement for high radiation areas.

{ The use of barricades and flashing lights is adequate protection for j individual high radiation areas to limit personnel accessibility.

5. l-4 I n s4. v 6 ( 4e.

                                                               -----{d?5.2-1 (kk rt)

JUSTIFICATION FOR DIFFERENCES TS 4 C A-5.o- ook 5/15/97

1 Insert for Enci. CA Sht 4 j C A 5.o-oo4k-5.6 This change adds to ITS 5.6.5 the reference to WCAP-12610. This WCAP was the basis for Amendment NO.110 which allowed the use of ZlRLO clad fuel rods and ZlRLO filler rods. ; I I l l l l 1 l j ( l 1 l

 .                                                                                                                                l t

i- i 4

   ,n-.                       . - , ,
                                                              ,       ,         ,                     -n.-r              -  --,e,

CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431. SECTION 5.0 Page 4 of 4 TECll SPEC CllANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMNCHE PEAK WOLF CREEK CALLAWAY 5.6-2 Deletes the EDG Report to reflect the recomendations of Yes 1 - not in CTS Yes Yes . Gl. 94-01. Removal of Accelerated Testing and Special Reporting Requirements for Emergency Diesel Generators." dated May 31, 1994 5.6 3 Revises report date in ITS 5.6.2. " Annual Radiological Yes - Consistent Yes - See LA 42/48 Yes Yes Environmental Operating Report ~ to be consistent with the with CTS and CTS. LA 78/77. 5.6 4 Revises Sections 5.6.1 and 5.6.3. Occupational Radiation Yes Yes Yes Yes Exposure Report" and " Radioactive Effluent Release Report." respectively, per NRC letter dated July 28. 1995. " Changes to Technical Specifications Resulting from 10 CFR 20 and 50.36a Changes" (From Christopher I. Grimes to Owners Groups Chairs). (TSTF-152) r 5.6-5 [ [ ] PORY lift sejtinm = referenced in PTLR section per Yes TR-5.o-oo3) Yes Yes Yes

             &'.l00 C',:N .-- 1.(15TP- 13%
           \

5.7-1 (RevisesHighRadiationAreatoincorporateconsistent Yes Yes Yes Yes

             , changes with 10 CFR 20.1601.         i fs5 e.i t 6 6 - 5 b s               ,g
                                                                                                                                          ~

5.7-2 Ch ges unaut' ri$br o "i dver ent' i t Hi -Ves--- Yes--. des---- Nes-- Ra fat n Are sectic to r let the *s si on as I s ted in RG .38, 5 ction .5 r gardi ph it barr ers Q5.1-(j r High Rad ation eas. This is c iste t w th tr veler ! TF-157.[If\5c rh (,6-Scj. 5.7 3 This change deletes the phrase ~or that cannot be No No No Yes continuously guarded" from the ITS for Callaway to make them consistent with the CTS. d) 5.'l -4 _fAlert (e IQ5.Z-(

5. (,- G In u t 6 6- S AMk 5-2 I l t

e a t t ifn i (f%MFrDCtent t'OtAD A D T Cent T Ant r an tDrf' 1 JI11

Insert for Encl. 6B Sht 4 C A- 5.o - oo4\ DC CP WC CA 5.6-7 This change adds as a reference to ITS 5.6.5, NO NO NO YES WCAP-12610, as an analytical method used to determine the core operating limits. 1 l l j l l l l l-l. J l

                                                          ,   .,              ,-s,.,                      , . . , . . . .          . , - -___  ,w-      <-,-

ADDITIONAL INFORMATION COVER SHEET

 ' ADDITIONAL INFORMATION NO: CA 5.0-005                   APPLICABILITY: CA REQUEST: ITS 5.5.6, remove the inserted redline text and add the following: "FSAR Chapter 16". This change should have been made by the RAI submittal to ITS Section 3.6, ULNRC-03853, TAC No. M98803, dated June 26,1098.

ATTACHED PAGES: Encl. 5A 5.0-10 a

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Procram (continued)

2. For' Iodine 131,7for Iodine 133,5for tritium,'and;for, a.l.1 radMn@lides2ngarticulate' form with:halfflives greater!than 8_ day.sF~Less than oriesual.to~a dosefatioH500 mrem /yr m tKany orgag

h. Limitations on the annual and quarterly air doses resulting from nobic gases released in gaseous effluents from each unit to areas beyond the rite boundary, conforming.to 10 CFR 50, Appendix I:

i. Limitations on the annual and quarterly doses to a member of the public from iodine 131, iodine 133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released fica cach unit to areas beyond the site boundary, conforming PS to 10 CFR 50, Appendix I; and

j. Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190:.:

5.5.5 Comoonent Cyclic or Transient Limit B.PS This program provides controis to track the FSAR, Section E3;9(N)'.~1217 "D,esign1TlainM6Ms*gh cyclic and transient occurrences to ensure that components are maintained within the design limits. 5.5.6 i'rc Strc;;cd Concicta Containment Tendon Surveillance Procram This program provides controls for monitoring any tendon degradation in pic- B PS stra;;d concrctc contain;;at;. including effectiveness of its corrosion protection medium, to ensure containment structural integrity. Tha program shall include baseline measurements prior to initial operations. The Tendon Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance

                      ,.m ,, _ ,.. with"thCST:0M,m_patit~1d[          psp;;feterRedWr.Tdf-'Regmtcry -
               %_;.-   y, m,- o,  .

_.., r :..

                             ,y-w u_-nm. m ar,<_.; onwvuiuwvis   ,,..

ww m.ww c.cr 4.ss, nww r:ra ivun r , n ,;,,- -

s. A m a.

The provisions of SR 3.0.2 and SR 3.0.3 are applicabla to the Tendon CA- 5-(,- oo t \ f Surveillance Program inspection frequencies. L FS% Cyb 16. / \.. _

                        ~                                                                       (continued)

MARK-UP OF WOG STS REV 1 (NUREG 1431) 5.0-10 5/15/97

l l ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: TR 5.0-003 APPLICABILITY: CA, CP, DC, WC i l REQUEST: ITS 5.6.6, " Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)", was revised to incorporate changes based upor "OG-67. WOG-67 has been approved by the TSTF and is designat 3 as TSTF-233. This traveler has been submitted l to the NRC and the latest traveler reports indicate that TSTF-233 has l been approved by the NRC. The attached pages reflect changes associated with WOG-67 being designated as TSTF-233. l ! ATTACHED PAGES: Encl. 5A Traveler Status Page Encl. 6A Page 4 Encl.68 Page 4 i i

l Industry Travelers Applicable to CTS Section 6.0/ITS 5.0 TRAVELER

STATUS DIFFERENCE # COMMENTS TSTF 9 Incorporated B -PS NRC Approved TSTF 37. Rev.1 Incorporated 5.6 2 DCPP only TSTF 52 Incorporated 5.5 4 [$'[Q3 ,\[ I g _y TSTF 65 Not Incorporated NA Not NRC approved as of !

traveler cut off date. . TSTF 106 Not Incorporated NA Retained CTS ; Rev. 1 ' TSTF 118 Incorporated 5.5 8 (UktAITroveh,lTR Rodoo6,) Not Incorporated TSTF 119 NA (%dtd:dCTS , cR$cM(TRs.o oo(T

TSTF-12 k) Not Incorporated

                                                                                                   %A tained CTS 4TR 5 o-cot T f/SJ    4 ik/   /     np6rpor/te,ef' [    //y h'               j             [$ 5.2-()

TSTF 152 Incorporated 5.6 4 dQ5.o Q (J$T(1fd / / IngdrpVated / / / y.7. !'gy ) l Q 5.2-/]

                 -* @b7.$gv./g                      Incorporated                    5.6 5               ]TR 5.o - oo3
                 ' ' Woc Q                           Incorporated                   5.5-13 7:"2C 2[

('T_STf-15 Incorporated 5.5 14 kQ S'5-kl Proposed Incorporated 5. 2.- 7., 5. 5 1, 5. '2- 3 Traveler Wp nyfni[dou)( [ l Acy'io { l'.[>[.[_~ 7 > SS-4 Jtem(147

                                                                                                                              -Y g
                                                                                        '                            ~

{TTTF- 258 _ - i l l I j MARK UP OF WOG STS REV 1 (NUREG 1431) 5/15/97

CHANGE NUMBER JUSTIFICATION 5.5 15 ITS Section 5.5.11c is being revised consistent with the CTS. The proposed changes specifies a time interval "of within 31 days after removal" in which a laboratory test of a sample obtained from each of the ESF systems charcoal adsorber must be tested for methyl iodide DenetratioIL - Inkrt GA- R h M[ 5.5-Uo 'I - 5.6 1 Not applicable to Callaway Plant. See Conversion Comparison Table (Enclosure 68). yg [ CA-S.o- oo3 } 5.6 2 This change deletes the EDG Report to reflect the recommendations of GL 94 01. " Removal of Accelerated Testing and Special Reporting Requirements for Emergency Diesel Generators," dated May 31, 1994. 5.6-3 This change revises the report date in Section 5.6.2, " Annual Radiological Environmental Operating Report" to be consistent with the CTS. 5.6 4 This change revises Sections 5.6.1 and 5.6.3, " Occupational Radiation Exposure Report" and " Radioactive Effluent Release Report," respectively, per NRC letter dated July 28,1995. " Changes to Technical Specifications Resulting from 10 CFR 20 and 50.36a Changes" (From Christopher I Grimes to Owners Groups Chairs). This is consistent ~ with traveler TSTF-152. , 5.6-5 lift settings are referenced in PTLR section per I n ~; TR 5.o-oo3} M 5.2.-l } N 1- - h . 6, . ', xn te.r t GA- 4b j o ist wit CF ttfe adia n sour e. InI<.rt 6A-4C. Q 5.2- 1~ 5.7 2 rev es "u uthor ed" t "ina

Thi ha rt "i the igh l iat' n Ar sect n to flec N s iti as tat in/ RG .8S tion .5 re rdin hysi bar iers or gh diaMon Qras. is i cons ent h tr eler STF- Insert 6A-5.7 3 ITS Section 5.7.3 is being revised consistent with the CTS. The proposed change deletes the phrase "or that cannot be continuously guarded" from the ITS to be consistent with the practices of the current TS which do not have this requirement for high radiation areas. The use of barricedes and flashing lights is adequate protection for individual high radiation areas to limit personnel accessibility. 5.I-4 I n 54. r 6 G A- 4 e. d2 5.2-1 i

 ~
                  < & ge D                                    C A-5 o- 00 JUSTIFICATION FOR DIFFERENCES - TS                  4                                          5/15/97

CONVERSION COMPARISON TABLE FOR DIFFERENCES FROM NUREG-1431. SECTION 5.0 Page 4 of 4 TECH SPEC CHANGE APPLICABILITY NUMBER DESCRIPTION DIABLO CANYON COMANCHE PEAK HOLF CREEK CALLAWAY 5.6-2 Deletes the EDG Report to reflect the reconnendations of Yes No - not in CTS Yes Yes GL 94 01. " Removal of Accelerated Testing and Special Reporting Requirements for Emergency Diesel Generators." dated Hay 31, 1994. 5.6-3 Revises report date in ITS 5.6.2. " Annual Radiological Yes - Consistent Yes - See LA 42/48 Yes Yes Environmental Operating Report" to be consistent with the with CTS and CTS. LA 78/77. 5.6-4 Revises Sections 5.6.1 and 5.6.3. " Occupational Radiation Yes Yes Yes Yes Exposure Report" and " Radioactive Effluent Release Report." respectively, per NRC letter dated July 28, 1995. " Changes to Technical Specifications Resulting from 10 CFR 20 and 50.26a Changes" (From Christopher I. Grimes to Owners Groups Chairs). (TSTF-152) r 5.6-5 [ ] PORY lift setti ay__ ara referenced in PTLR secti Yes TR-5.o-oo3) Yes Yes Yes A= c. m.. :.ctsw- 13A D 7 5.7-1 fRevisesHighRadiationAreatoincorporateconsistent] Yes Y'S Y'S Y'S changes with 10 CFR 20.1601._ InSevt 6 6-5b s - t G 5.1- h 5.7-2 'Ch .s unaut ri$d" o "i dyer ent" t Hi -Ves--- Vee-- .b wrs-- Ra fat n Are sectio to r .lec the 's si on as

                                                              .38. 5 ction .5 r ardi s ated in RG frHtph Rad ation                       eas. This is ic barr ers iste w th tr veler                                                                                             fQ51-(}             !

TF-157 7 If\Se r6 66-Scj-5.7-3 This change deletes the phrase "or that cannot be No No No Yes continuously guarded" from the ITS for Callaway to make them consistent with the CTS. 5.1--4 fAfert G h 5.7. - (

5. c -(, Inu,t66-SA M 52 I(

r*OnfurD C T nM ('nuDADTCnu TADIC 619 fntf' 1A11 e si t in7

i ADDITIONAL INFORMATION COVER SHEET ADDITIONAL INFORMATION NO: TR 5.0-006 APPLICABILITY: CA, CP, DC, WC REQUEST: Revise the Traveler Status Sheet to reflect the latest status and revisions of the following travelers: TSTF-118 - NRC Approved j TSTF-119 - NRC Rejected TSTF-120, Rev.1 TSTF-152 - NRC Approved ATTACHED PAGES: Encl.5A Traveler Status Page 1 l I (

i Industry Trevelers Applicable to CTS Section 6.0/ITS 5.) TRAVELER # STATl)5 DIFFERENCE # COMMENTS TSTF-9 Incorporated B PS NRC Approved TSTF 37, Rev.1 Incorporated 5.6-2 DCPP only TSTF-52 Incorporated 5.5 4 f('[f h.#[' 3 3 3.i_g j TSTF 65 Not Incorporated NA nlot NRC approved as of traveler cut off date. TSTF 106 Not Incorporated NA Retained CTS Rev. 1 TSTF 118 Incorporated 5.5-8 (5kC. AiTrovh "ITR Rodood l TSTF 119 Not Incorporated NA (]5tci=d:T5 )JRC R<icchTR5.o.odh l TSTF 12 h Not Incorporated % Retained CTS l TR 5.o-00t. T h/S}f1/1 / [ /Inp6rpor/teg' [ //yh'j [ $ 5.2-() TSTF 152 Incorporated 5.6 4 k/N TK 5.o. oof.1 y (J$TE'15(j f l / Inq6roWated / / / ,E(.7-[gy lQ5*L-I}

            % F 7,t g r/./'g          Incorporated              5.6 5            jn 5.o - co3
                  'WOGQ               Incorporated             5.5 13 (TSTf -13 7 9CC S h Incorporated             5.5-14                 kQ 6 I~ k_

Proposed Incorporated 5. 7.. z., 5. 5 1, 5'. 2- 3 Wp nyfni-grouf Acfio s Traveler 5 ll' 3 p_;

2. , s.s-l' / tem (147( ( l

                                                                     '                       ~
                  "TETF- 258     _

g i i i MARK UP OF WOG STS REV 1 (NUREG-1431) 5/15/97 -}}

ML20154A290

Text

6.0 DESCRIPTION

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