Pages: 2.0-1 I - NRC

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ATTACHMENT 3'to TXX 99113

ITS TECHNICAL SPECIFICATION MARKUP

Pages: 2.0-12.0 22.0 3 i

B 2.0 1 |B 2.0 2B 2.0-3B 2.0 4

3.3'21 !

3.4 1 I3.4 2 -!

3.4 3B 3.4 1B 3.4 2B 3.4 3

5.0 315.0 325.0 33

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9906020125 99052445jDR. ADOCK'O

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SLs2.0

2.0 SAFETY LIMITS (SLs)

2.1 SLs

2.1.1 Reactor Core SLs

|n MODCC i cad 2, th; ;;mb| action of T|iERMAL POWER, RcccistOco' cat Cy:;tcm (ROC) h|ghcot | cop cvcccg; icrapercturc, cad prc:;;uriccrpic:;;;urc chci! not ex;ccd the CL; cpc;|f|cd in figurc 2.1.1 1.2;1;14 in MODES;1^and;2,5the'departoreifrom nucleate boill.ng ratio

(DNBR).;shall.be maintainedXthe;95/95 DNB criterion for theDNB'correlatiqq(sLspecifiedjn SectlpA5 6,5

2.1;1.2 In(MODESTand'2,'the peaKfuel centerline, temperature shallbeimaigta.ined14700*ff

2.1.2 RCS Pressure SL

in MODES 1,2,3,4, and 5, the RCS pressure shall be maintaineds 2735 psig.

2.2 SL Violations

2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour.;

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within1 hour.

|2.2.2.2 In MODE 3,4, or 5, restore c.ompliance within 5 minutes.

COMANCHE PEAK - UNITS 1 AND 2 2.0-1 Amendment No. 64

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Figure 2.1.1-1 (page 1 of 2)Reactor Core Safety Limits (Unit 1)

[THIS FIGURE AND PAGE HAVE BEEN DELETED.]

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Figure 2.1.1-1 (page 2 of 2)Reactor Core Safety Limits (Unit 2)

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Reactor Core SLsB 2.1.1'

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.B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs i

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BASES ~

BACKGROUND. GDC 10 (Ref.1) requires that specified acceptable fuel design limits arenot exceeded during steady state operation, normal operational !

transients, and anticipated operational occurrences (AOOs). This isaccomplished by having a departure from nucleate boiling (DNB) design .

Ibasis, which corresponds to a 95% probability at a 95% confidence level(the 95/95 DNB criterion) that DNB will not occur and by requiring thatfuel centerline temperature stays below the melting temperature.

The restrictions of this SL prevent overheating of the fuel and cladding,as well as possible cladding perforation, that would result in the releaseof fission products to the reactor coolant. Overheating of the fuelisprevented by maintaining the steady state peak linear heat rate (LHR)below the level at which fuel centerline melting occurs. Overheating ofthe fuel cladding is prevented by restricting fuel operation to within thenucleate boiling regime, where the heat transfer coefficient is large andthe cladding surface temperature is slightly above the coolant saturationtemperature.

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Fuel centerline melting occurs when the local LHR, or power peaking, ina region of the fuelis high enough to cause the fuel centerlinetemperature to reach the melting point of the fuel. Expansion of thepellet upon centerline melting may cause the pellet to stress the claddingto the point of failure, allowing an uncontrolled release of activity to thereactor coolant.

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Operation above the boundary of the nucleate boiling regime could resultin excessive cladding temperature because of the onset of DNB and theresultant sharp reduction in heat transfer coefficient. Inside the steamfilm, high cladding temperatures are reached, and a cladding water(zirconium water) reaction may take place. This chemical reactionresults in oxidation of the fuel cladding to a structurally weaker form.This weaker form may lose its integrity, resulting in an uncontrolledrelease of activity to the reactor coolant.

=The proper functioning of the Reactor Protection System (RPS) and jsteam generator safety valves prevents violation of the reactor core SLs.

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COMANCHE PEAK- UNITS 1 AND 2 B 2.0-1 Amendment No. 64

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| Reactor Core SLsB 2.1.1

BASES (continued)

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| APPLICABLE The fuel cladding must not sustain damage as a result of normalSAFETY operation and AOOs. The reactor cora SLs are established to precludeANALYSES violation of the following fuel design criteria:

a. 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 notexperience DNB; and

b. The hot fuel pellet in the core must not experience centerline fuelmelting.

The Reactor Trip System Allowable Values in Table 3.3.1-1, in| combination with all the LCOs, are designed to prevent any anticipated

combination of transient conditions for Reactor Coolant System (RCS)temperature, pressure, RCS flow, Al, and THERMAL POWER level thatwould result in a departure from nucleate boiling ratio (DNBR) of lessthan the DNBR limit and preclude the existence of flow instabilities.

Protection for these reactor core SLs is provided by the appropriate. operation;of the;RPS:and the steam generator safety valves; emHhe! fc||cv.ing automat |c rcactor tr|p funct|cas:

C. M|gh pic;;ur|zcr pic;;;ure trip;

b. LcVi pic;;ur|zcr pic;;ure trip;

c. OVcrtCmpcraturC N-16 trip;

d. Ovcip;;;;r N-1S tr|p; cnd

i c. Ic?;c Rong; Ncutron I|ux |||gh tr|p.

The ||mitat|cn that the averagc cathc|py in the hot |c; bc |c;; than orcqualis the cath;|py of saturated || quid is not a cccc protcction ||mit, but I

cn;urc; that thccc Orc no vc|d; |n the het |c; thct ;cu|d Offcci th; N-10;|gn;| den;;ty.

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COMANCHE PEAK - UNITS 1 AND 2 B 2.0-2 Amendment No. 64 j

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| Reactor Core SLsB 2.1.1

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APPLICABLE- .The SLs represent a design requirement for establishing the RPSSAFETY Allowable Values identified previously. LCO 3.4.1,"RCS Pressure,ANALYSES Temperature, and Flow Departure from Nucleate Bolling (DNB) Limits,"

(continued) and the assumed initial conditions of the safety analyses (as indicated in! the FSAR, Ref. 2) provide more restrictive limits to ensure that the SLs

are not exceeded.

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| SAFETY LIMITS The curv;; p.;v d:d in "|gur; 2.1.1_1 ;h;;; th; |c;; ;f p; n'.;_cf| ...-n u..m.. . - . A m m.i., r n. , m, , e e. r_ _ _ _ _ . . _~ .., ....J ..._____a.....r..... .L.._t_..

_ _ _______A. . -. w. . . . . . . .. . . . . .,

' veh!;h th; 0;|cu|;t-d ONO",i; not |:= then the de-|gn ONO", v;|ac, f;;|I ;;nt r| n; i;mp;retur; .; mein; b;| eve in-|'|ng, 'h; ;v;;;;; enth-|py inI theMt |:g !; |c= th;n er ;qu;| to th; enth;|py ;f ;;tur;';d ||qu|d, or th;

_...a _ . . _ t ! A. . f_. .!ALf_A L ._ s u s i vi te .. s a u s..J r%ktrtP,tE! ? A_ J _ f" _ _ L..AL_ _____f_At__

.ny t q ua.si ty is .. s t u r v i , ti v 7 sv u. w a iws ..v a . s. s s.v u .

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The curvcs crc bcscd on enthc|py risc het chcnnc| fcstcr !!rnits prov dcdin th; OOL" ;nd ; Cy;|; ;pe;|f|; repr;;;n'et|v; exi;| p;;;;r ;h;p;. ^n, ,

O!! t'!:^00 !: !" d0d f0r 0" M0 0000 !" % ot 7;du;;d p;;;;r b;;;d onthc equ;t|cn g ven |n 'h; OOL",.

7" L ._ A L ._ .. t._ %s r. s. se A L ._ r- ,r* M 1_ . .!AL?- AL _ef 1. Lt_L__AL__ t ! __ ! A __ ..t_AJ ..L__ AE II wh te surgis. nsuuss ni g rru a rst r us.us ts r s w re .. s n. s vi s ti v.

! "-!!: Of !5; F,(^.|} fu .;uen of th; cvertemperature re;;ter tr:p. '.".'he nA. L. ._,A.r. * M, f ._

__A &_.! A L ? _A. L. ._A.._f______ A.L..._m ,"r4 .....A. __ AL_Ar _ ff _ _

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. . . . . . . . . . . . . ......m, w ... ,,z., ,

overt;mper;tur; r;;; tor tr:p; w;||| .;du;; 'h; ;;tp;|nt; to prev |d;,

l pret;;t en ;;n;;;t;nt ;;;th th; r;;;ter ;;;; OL; (",;f. 0 ).The reacto.r; core:SLs:are:establi.shedjto;precludellolationMthe i

following f.u_el,de_ sign c_riter_ia; Ij_

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al There must b.e at least a 95Wprobability~at~a 95% ;

confidence. level. (the 95/95, DNB;; criterion) that;.the;hotifuel I

rod;injheicore does not_ experience Df9B;;and,

b; There must be:at least a 95Wprobability:st'a 9591iconfidenoe level that the hot fuel pellet in ti)e., core dosa;notexperience oente.rline fueltneitingj

The reactor _ core:SLs~are:used..toidefine;theivarious|RPSilunctionsisualil' that the:above: criteria;are~. satisfied during steady; state operationgnormaloperational transients; Mand anticipated; operational coeurrences'.(AOOs)gTo. ensure,that.the RPS precludes theylolation;of<the'above criterialadditional criteria:are applied,to the:Overtemperature N-16 reactor. tripfun.ctionsghat,lsfit must be'. demonstrated that the average enthalpy inthe hot leg is less than or, equal to the saturationienthalpy_and that, tilecore . exit qualityjs within;the lirnits defined by the;DNB, correlation;]

COMANCHE PEAK - UNITS 1 AND 2 B 2.0-3 Amendment No. 64

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Reactor Core SLsB 2.1.1

BASES (continued) .

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Appropriate; functioning of the RPS and the steem safetyvalves, ensure that foryariationsJn theJHERMAL _ _ ,RCSPressureRRCS| average temperaturet.RCS flow talerandMthat thereactoteprei8t.siwill be sellsfied;during steedKatste_oporptiongnonna1

. operational transients;and3OOsj

APPLICABILITY . SL 2.1.1 only applies in MODES 1 and 2 because these are the onlyMODES in which the reactor is critical. Automatic protection functionsare required to be OPERABLE during MODES 1 and 2 to ensureoperation within the reactor core SLs. The steam generator safetyvalves or automatic protection actions serve to prevent RCS heatup tothe reactor core SL conditions or to initiate a reactor trip function, whichforces the unit into MODE 3. Allowable Values for the reactor tripfunctions are specified in LCO 3.3.1, " Reactor Trip System (RTS)Instrumentation." In MODES 3,4,5, and 6, Applicability is not requiredsince the reactor is not generating significant THERMAL POWER.

SAFETY LIMIT The following SL violation responses are applicable to the reactor coreVIOLATIONS SLs. If SL 2.1.1 is violated, the requirement to go to MODE 3 places the

unit in a MODE in which this SL is not applicable.

The allowed Completion Time of 1 hour recognizes the importance ofbringing the unit to a MODE of operation where this SL is not applicable,and reduces the probability of fuel damage.

Per 10CFR50.36, if a Safety Limit is violated, operations must not beresumed until authorized by the Commission.

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REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.

2. FSAR, Chapter 7.|

3. * Power Distribution Control Analysis and Overtemperature N-16 jand Overpower N-16 Trip Setpoint Methodology," RXE-90- !006-P-A, June 1994. '

COMANCHE PEAK'- UNITS 1 AND 2 B 2.0-4 Amendment No. 64

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RTS Instrumentation3.3.1

hatt.11 Overtemoerature N 16

The Overtemperature N 16 Function Allowable Value shall not exceed the followingsetpoint by more than 1.72% of span for Unit 1, or 2.82% of span for Unit 2.

(1 + T,s) T -T" + K,(P - P')- f,(A q)Q,,,,,u, = K,- K, c c

Where:

Q,,,,,oi = Overtemperature N-16 trip setpoint,

K, 4-%O *s,=

K,-

0.01""' f:: Uni 29-044Bi'fF f; Uni'.1=

K = 040080;*1psig-fee-Un++3,

. . . . . . , , , . . , , . . . . . .

Tc Cold leg temperaturo=

be{e,rgce T,c,_at, RATED THERMAL POWER, C'E* e8.

- E!E+ E EidP = Measured pressurizer pressure, psig

P' a E0863 psig (Nominal RCS operating pressure)

the Laplace transform operator, sec-',s =

T, ,T, = Time constants utilized in lead-lag controlier for T.,T, a 49j sec, and 7, s 82 sec

f,(Aq) =

G-09f((q, q.) + 663%)when (q,- q ) s 662% RTPUn'' " 0% when-662% RTP < (q,- q ) < +-6-03% RTP

B-41((qi- go) 64f%)when (q,- q.) a +-6-92% RTP1

0,404%- qd * SS%' " hen (;c qd SS% P "L'n9 2: C a' : hen 95% PT ' (;,--qJ ' 5.1 % P

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; 3,3Sy %--qd S.*%' -h: . (qc--qJ ; - 5. ' % P"|

| Note 2. Not Used.

Les specified irt the1 COLL 3


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RCS Pr:ssura, Temperatura, and Flow DNB Limits3.4.1

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3,4 REACTOR COOLANT SYSTEM (RCS)

3.4.1 RCS Pressure. Temperature, and Flow Departure from Nucleate Boiling(DNB) Limits

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LC0 3.4.1 RCS DNB parameters for pressurizer pressure, RCSaverage temperature, and RCS total flow rate shall be

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within the limits specified below:'

a. Pressurizer pressure 2Ethe21mitispecified';jnithelCOLR2210 p;ig:

|b. RCS average temperature sLthel]init;;specifjedlinithe

COLR259B4: and

c. RCS total flow rate 2 403.400 3897700 gpmandi,thelimitIspecifiedI1Ethe[COLRifer Unit 1

2 400,000 gp;;; for Unit 2.

APPLICABILITY: MODE 1

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Pressurizer pressure limit does not apply during:

a. THERMAL POWER ramp > 5% RTP per minute: or.

b. THERMAL POWER step > 10% RTP.............................................................

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ACTIONS!

| CONDITION REQUIRED ACTION COMPLETION TIME! _

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| A. One or more RCS DNB A.1 Restore RCS DNB 2 hours| parameters not within parameter (s) to

limits. within limit.

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i'| COMANCHE PEAK - UNITS 1 AND 2 3.4-1 Amendment No. 64

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l' RCS Pr:ssura, Temperatura, and Flow DNB LimitsI- 3.4.1

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ACTIONS (continued) .

CONDITION REQUIRED ACTION COMPLETION TIME

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L B. - - NOTE - B.1 Maintain THERMAL Immediately!

POWER less than 85%--

|- Only applicable prior RTP.| to exceeding 85% RTP

after a refuelingoutage.......................

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Measured RCS Flow not'

within limits.

C. Required Action and C.1 Be in MODE 2. 6 hoursassociated CompletionTime not met.

SURVEILLANCE REQUIREMENTS

SURVEILLANCE FREQUENCY

SR 3.4.1.1Verifypressurizerpressureis&theGimit 12 hoursspecifiedi1Mthe;COLRt 2210 p;ig.

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SR 3.4.1.2VerifyRCSaveragetemperatureis$ithellfait; 12 hoursspecifjedfjnJthe;COLR2 s 502"f. i

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SURVEILLANCE FREQUENCY

SR 3.4.1.3 Verify RCS total flow rate isS&389@0;4ndh2 12 hourst hem i si.tispeci fi.edlj n S the'C0,LRs

a 400,400 gp;r. for Un:t i. 400,000 gpm. for Un|t 2.

SR 3.4.1.4 ----NOTE------ ------ ----------.......- .

Not requiEd to be performed until after exceeding 85% RTPafter each refueling 6utage............... .............................. ...... ..... ...

18 monthsVerify by precision heat balance that RCS total flowrate is'k 389200 andAthejimit specified in'the_COLR:~ T400,400 gp,T, for Unit i

< 400,000 gp,T,for U nit 2.

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RCS Prassura, Tempsratura, and Flow DNB LimitsB 3.4.1

B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)Limits

BASES

BACKGROUND ' These Bases address requirements for maintaining RCS pressure,temperature, and flow rate within limits assumed in the safetyanalyses. The safety analyses (Ref.1) of normal operating conditionsand anticipated operational occurrences assume initial conditionswithin the normal steady state envelope. The limits placed on RCS )pressure, temperature, and flow rate ensure that the minimumdeparture from nucleate boiling ratio (DNBR) will be met for each ofthe transients analyzed.

The RCS pressure limit is consistent with operation within the nominaloperational envelope. Pressurizer pressure indications are averaged tocome up with a value for comparison to the limit. A lower pressure willcause the reactor core to approach DNB limits.

The RCS coolant average _ temperature limit is consistent with full poweroperation within the nominal operational envelope. Indications of -temperature are averaged to determine a value for comparison to the limit. I

IA higher average temperature will cause the core to approach DNB limits.

iThe RCS flow rate normally remains constant during an operational fuelcycle with all pumps running. The minimum RCS flow limit corresponds tothat assumed for DNB analyses and includes an allowance of 1.8% flow formeasurement uncertainties. Flow rate indications from the plant computeror RCS flow rate indicators are averaged to come up with a value forcomparison to the limit during shiftly surveillances. A lower RCS flow willcause the DNB limits to be approached. After each refueling, the elbow tapdifferential pressure transmitters are normalized to the precision RCS flowmeasurement. The uncertainty associated with the RCS flow measurement(1.8%)is based on the use of the feedwater venturis and precisioninstrumentation which has been calibrated within 90 days of performing thecalorimetric flow measurement.

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APPLICABLE Operation for significant pariods of tims outsida tha limits on RCS flow,SAFETY pressurizer pressure and average temperature increases the likelihoodANALYSES of a fuel cladding failure if a DNB limited event were to occur.

The requirements of this LCO represent the initial conditions for DNBlimited transients analyzed in the plant safety analyses (Ref.1). Thesafety analyses have shown that transients initiated from the limits ofthis LCO will result in meeting the DNBR criterion. This is theacceptance limit for the RCS DNB parameters. Changes to the unitthat could impact these parameters must be assessed for their impacton the DNBR criterion. The transients analyzed for include loss ofcoolant flow events and dropped rod events. A key assumption for theanalysis of these events is that the core power distribution is within thelimits of LCO 3.1.7, " Control Bank Insertion Limits"; LCO 3.2.3, " AXIALFLUX DIFFERENCE (AFD)"; and LCO 3.2.4," QUADRANT POWERTILT RATIO (QPTR)."

The pressurizer pressure limit of 2210 p;4 and the RCS averagetemperature limit specifiedjn the COLR ci 502*I correspond to $aanalyticallimits ef-2205 p;|;;:nd 504.7 I(for Un|t i,505.2'r for Un|tE) used in the safety analyses, with allowance for measurementuncertainty. These uncertainties are based on the use of controlboard indications. J

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The RCS DNB parameters satisfy Criterion 2 of 10CFR50.36(c)(2)(li).

I'LCO This LCO specifies limits on the monitored process variables-

pressurizer pressure, RCS average temperature, and RCS total flowrate - to ensure the core operates within the limits assumed in thesafety analyses. Theseyariables'are~ considered _in theLCOLR toprovide operating and, analysis flexibihty,from.cycleito; cycle. . However;the minimum;RCS| flow? based on maximum 3palyzed;steegt generatortube plugging, isjetained in:the;TSACO| Operating within these limitswill result in meeting the DNBR criterion in the event of a DNB limitedtransient.

RCS total flow rate contains a measurement error of 1.0"' based onperforming a precision heat balance and using the result to normalizethe RCS flow rate indicators. Potential fouling of the feedwaterventuri, which might not be detected, could bias the result from theprecision heat balance in a nonconservative manner.

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RCS Pressure, Temperature, and Flow DNB Limits |B 3.4.1 1

BASES

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LCO Any fouling that might significantly bias the flow rate measurement can |

(continued) be detected by monitoring and trending various plant performance be |

parameters. If detected, either the effect of the fouling shall bequantified and compensated for in the RCS flow rate measurement orthe venturi shall be cleaned to eliminate the fouling. CPSES also usesthe Transit Time Flow Meter (TTFM) to measure the volumetric RCGhot leg flow rate. The use of the TTFM results in an RCS flowmeasurement which is more accurate and less sensitive to RCS fluidconditions than other methods.

The L-GO numerical values for pressure, temperature, and flow ratespecified in;the C_OLf3 have been adjusted for instrument error.

APPLICABILITY In MODE 1, the limits on pressurizer pressure, RCS coolant averagetemperature, and RCS flow rate must be maintained during steadystate operation in order to ensure DNBR criteria will be met in theevent of an unplanned loss of forced coolant flow or other DNB limitedtransient. In all other MODES, the power level is low enough that DNBis not a concern. j

A Note has been added to indicate the limit on pressurizer pressure isnot applicable during short term operational transients such as aTHERMAL POWER ramp increase > 5% RTP per minute or aTHERMAL POWER step increase > 10% RTP. These conditionsrepresent short term perturbations where actions to control pressurevariations might be counterproductive. Also, since they representtransients initiated from power levels < 100% RTP, an increasedDNBR margin exists to offset the temporary pressure variations.

Ar,cther act of ||mits or, DNS rc';ted paramcicr; The_.DNBR limit isprovided in SL 2.1.1, " Reactor Core SLs." Tho;; ||m ts Th.e_ conditions

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which define the DNBR limit are less restrictive than the limits of thisLCO, b0t violatiori of 5 Safety Limit (SL) merits a stricter, more severeRequired Action. Should a violation of this LCO occur, the operatormust check whether or not an SL may have been exceeded.

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COMANCHE PEAK - UNITS 1 AND 2 8 3.4-3 Amendment No. 64

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'5.6.5 CORE OPERATING LIMITS REPORT (COLR)

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Core operating limits shall be established prior to each reload cycle, ora.prior to any remaining portion of a reload cycle, and shall bedocumented in the COLR for the following:

1) Moderator temperature coefficient limits for Specification 3.1.3,

2) Shutdown Rod insertion Limit for Specification 3.1.5,

3) Control Rod Insertion Limits for Specification 3.1.6,

4) AXIAL FLUX DIFFERENCE Limits and target band forSpecification 3.2.3,

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5) Heat Flux Hot Channel Factor, K(Z), W(Z), FoRTP, and the Fa (Z)C

allowances for Specification 3.2.1,!

6) Nuclear Enthalpy Rise Hot Channel Factor Limit and the Power jFactor Multiplier for Specification 3.2.2. !

7) SHUTDOWN MARGIN values in Specifications 3.1.1,3.1.4,3.1.5, 3.1.6 and 3.1.8.'

8) Refueling Boron Concentration limits in Specification 3.9.1.

9) Overtemperature;Nji61 Trip,Setpoint.irLSpecificationi3;311

10) Reactor Coolant . System;pressureftemperature;ian.d. flow irlSpecification _3;411

b. The analytical methods used to determine the core operating limhsshall be those previously reviewed and approved by the NRC,specifically those described 'in the following documents:

1) WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETYEVALUATION METHODOLOGY," July 1985 (W Proprietary). i

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2) WCAP-8385," POWER DISTRIBUTION CONTROL AND LOAD |,

| FOLLOWING PROCEDURES - TOPICAL REPORT," September j'

1974 (E Proprietary).

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R:: porting Requirements5.6

5.6 Reporting Requirements;

I 5.6.5 CORE OPERATING LIMITS REPORT (continued)

3) T. M. Anderson To K. Kniel(Chief of Core Performance Branch,NRC) January 31,1980--Attachment: Operation and SafetyAnalysis Aspects of an Improved Load Follow Package.

4) NUREG-0800, Standard Review Plan, U.S. Nuclear RegulatoryCommission, Section 4.3, Nuclear Design, July 1981. BranchTechnical Position CPB 4.3-1, Westinghouse Constant AxialOffset Control (CAOC), Rev. 2, July 1981.

5) WCAP-10216-P-A, Revision 1 A, " RELAXATION OF CONSTANTAXlAL OFFSET CONTROL Fa SURVEILLANCE TECHNICALSPECIFICATION," February 1994 (W Proprietary).

6) WCAP-10079-P-A,"NOTRUMP, A NODAL TRANSIENT SMALLBREAK AND GENERAL NETWORK CODE," August 1985,(WProprietary).

7) WCAP-10054-P-A, " WESTINGHOUSE SMALL BREAK ECCSEVALUATION MODEL USING THE NOTRUMP CODE", August1985, (W Proprietary).

8) WCAP-11145-P-A, " WESTINGHOUSE SMALL BREAK LOCAECCS EVALUATION MODEL GENERIC STUDY WITH THENOTRUMP CODE", October 1986, (W Proprietary).

9) RXE-90-006-P, " Power Distribution Control Analysis andOvertemperature N-16 and Overpower N-16 Trip SetpointMethodology, " February 1991,

10) RXE-88-102-P,"TUE-1 Departure from Nucleate BoilingCorrelation", January 1989.

11) RXE-88-102-P, Sup.1, "TUE-1 DNB Correlation - Supplement 1",December 1990.

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R: porting R:quirements5.6

5.6 Reporting Requirements (continued)

|5.6.5 CORE OPERATING LIMITS REPORT (continued) |

|12) RXE-89-002,"VIPRE-01 Core Thermal-Hydraulic Analysis

Methods for Comanche Peak Steam Electric Station LicensingApplications", June 1989.

13) RXE-91-001," Transient Analysis Methods for Comanche PeakSteam Electric Station Licensing Applications", February 1991.

14) RXE-91-002," Reactivity Anomaly Events Methodology", May1991.

15)' RXE-90-007,"Large Break Loss of Coolant Accident Analysis'

Methodology", December 1990.

16) TXX-88306," Steam Generator Tube Rupture Analysis", March j15,1988.

17) RXE-91-005," Methodology for Reactor Core Response toSteamline Break Events," May,1991.

18) RXE-94-001-A," Safety Analysis of Postulated inadvertent BoronDilution Event in Modes 3,4, and 5," February 1994.

19) RXE-95-001-P,"Small Break Loss of Coolant Accident Analysis,

Methodology," December 1995.

c. The core operating limits shall be determined such that all 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 accident analysis limits) ofthe 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.

(continued)

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COMANCHE PEAK- UNITS 1 AND 2 5.0-33 Amendment No. 64

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Attachment 2 to TXX-99113' Page 8 of 8 '

approved methodologies. All accident analyses, performed in accordance with thesemethodologies, must meet the applicable, NRC-approved limits of the safety analysis.These changes are essentially administrative and do not change the type or quantity ofeffluents released offsite, nor will these changes increase individual or cumulativeoccupational radiation exposure. Based on the preceding evaluation, these changes donot involve a significant hazards consideration.

TXU Electric has evaluated the proposed changes and has determined that the changesdo not involve (i) a significant hazards consideration, (ii) a significant change in the typesor significant increase in the amounts of any effluent that may be released offsite, or (iii)a significant increase in individual or cumulative occupational radiation exposure.Accordingly, the proposed changes meet the eligibility criterion for categorical exclusionset forth in 10CFR51.22(c)(9). Therefore, pursuant to 10CFR51.22(b), an environmentalassessment of the proposed change is not required.

VI. REFERENCES

1. Generic Letter 88-16 " Guidance for Technical Specification Changes for Cycle-Specific Parameter Limits," October 4,1988

2. WCAP 14483,'" Generic Methodology for Expanded Core Operating Limits Report,"November 1995

3. NRC letter from Thomas H. Essig to Mr. Andrew Drake, (Westinghouse Owner'sGroup), dated January 19,1999

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