ML20195B677

From kanterella
Jump to navigation Jump to search
Proposed Tech Specs Relocating Parameter Limits Allows for Available Operating & Analytical Margins to Be Used in Most Efficient Manner
ML20195B677
Person / Time
Site: Comanche Peak  Luminant icon.png
Issue date: 05/24/1999
From:
TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:
Shared Package
ML20195B663 List:
References
NUDOCS 9906020125
Download: ML20195B677 (19)
v • d • e


Text

~

F. 1 q .,

l L

l l

1 ATTACHMENT 3'to TXX 99113 ITS TECHNICAL SPECIFICATION MARKUP Pages: 2.0-1 2.0 2 2.0 3 i B 2.0 1 l B 2.0 2 B 2.0-3 B 2.0 4 3.3'21  !

3.4 1 I 3.4 2 -!

3.4 3 B 3.4 1 B 3.4 2 B 3.4 3 5.0 31 5.0 32 5.0 33 i

j h

9906020125 99052445 DR. ADOCK'O j

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs ln MODCC i cad 2, th; ;;mbl action of TliERMAL POWER, Rcccist Oco' cat Cy:;tcm (ROC) hlghcot l cop cvcccg; icrapercturc, cad prc:;;uriccr pic:;;;urc chci! not ex;ccd the CL; cpc;lflcd 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 the DNB'correlatiqq(sLspecifiedjn SectlpA5 6,5 2.1;1.2 In(MODESTand'2,'the peaKfuel centerline, temperature shall beimaigta.ined14700*ff 2.1.2 RCS Pressure SL in MODES 1,2,3,4, and 5, the RCS pressure shall be maintained s 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 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

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 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

l 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

7 u ,

L SLs 2.0 Figure 2.1.1-1 (page 1 of 2)

Reactor Core Safety Limits (Unit 1)

[THIS FIGURE AND PAGE HAVE BEEN DELETED.]

I i

j l

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

)

f  !

J SLs 2.0

(

Figure 2.1.1-1 (page 2 of 2)

Reactor Core Safety Limits (Unit 2)

[THIS FIGURE AND PAGE HAVE BEEN DELETED.]

J COMANCHE PEAK - UNITS 1 AND 2 2.0-3 Amendment No. 64 4

n

]

Reactor Core SLs B 2.1.1

.B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs i i

BASES ~

BACKGROUND. GDC 10 (Ref.1) requires that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational  !

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

basis, which corresponds to a 95% probability at a 95% confidence level I (the 95/95 DNB criterion) that DNB will not occur and by requiring that fuel 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 release of fission products to the reactor coolant. Overheating of the fuelis prevented by maintaining the steady state peak linear heat rate (LHR) below the level at which fuel centerline melting occurs. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime, where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. 3 i

Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuelis high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant.

Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of DNB and the resultant sharp reduction in heat transfer coefficient. Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form.

This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

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

l 1

(continued) I i

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

a

[

l l

l Reactor Core SLs B 2.1.1 BASES (continued) 1 1

l APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY operation and AOOs. The reactor cora SLs are established to preclude ANALYSES 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 not experience DNB; and
b. The hot fuel pellet in the core must not experience centerline fuel melting.

The Reactor Trip System Allowable Values in Table 3.3.1-1, in l 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 that would result in a departure from nucleate boiling ratio (DNBR) of less than 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

! fcllcv.ing automat lc rcactor trlp functlcas:

C. Mlgh pic;;urlzcr pic;;;ure trip;

b. LcVi pic;;urlzcr pic;;ure trip;
c. OVcrtCmpcraturC N-16 trip;
d. Ovcip;;;;r N-1S trlp; cnd i c. Ic?;c Rong; Ncutron Ilux lllgh trlp.

The llmitatlcn that the averagc cathclpy in the hot lc; bc lc;; than or cqualis the cath;lpy of saturated ll quid is not a cccc protcction llmit, but I cn;urc; that thccc Orc no vcld; ln the het lc; thct ;culd Offcci th; N-10

lgn;l den;;ty.

l (continued) l l

l l

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

4

p I

l l Reactor Core SLs B 2.1.1 l

l l' BASES l

l-l APPLICABLE- .The SLs represent a design requirement for establishing the RPS SAFETY 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.

l l SAFETY LIMITS The curv;; p.;v d:d in "lgur; 2.1.1_1 ;h;;; th; lc;; ;f p; n'.;_cf

e. r_ _ _ _ _ . . _~ .., ....J ..._____a.....r.....

l _ _ _______A

, ...-n u..m.. . - . A . m .m.i.,

-. r n. , m, , ew. . . . . . . .. ..L.._t_..

veh!;h th; 0;lcul;t-d ONO",i; not l:= then the de-lgn ONO", v;lac, f;;l I

nt rl n; i;mp;retur; .; mein; b;l eve in-l'lng, 'h; ;v;;;;; enth-lpy in I theMt l
g !;f_.lc= th;n er ;qu;l to th; enth;lpy ;f ;;tur;';d llquld, or th;

_...a .!ALf_ E! L..AL_

_ . . _ t ! A. .  ? A_ J _ f" _ _ _____f_At__

.ny t q ua.si ty is .. s t u r v i , A ti L v ._ s u s i vi te .. s a u s..J r%ktrtP,t ..v 7 sv u. w a iws a . s. s s.v u .

The curvcs crc bcscd on enthclpy risc het chcnncl fcstcr !!rnits prov dcd in th; OOL" ;nd ; Cy;l; ;pe;lfl; repr;;;n'etlv; exi;l p;;;;r ;h;p;. ^n , ,

O!! t'!:^00 !: !" d0d f0r 0" M0 0000 !" % ot 7;du;;d p;;;;r b;;;d on thc equ;tlcn g ven ln 'h; OOL",.

ef 1. Lt_L__AL__ ..L__

A L ._Ar-s ,r*

t ! __ ! A __ ..t_AJ 7"E II L ._ wh te surgis. nsuuss ni g A L ._ rru a rst .. t._r %s r. s. se us.us ts r w reM..1_s n.

. .!AL?- ALv._

s vi s ti

! "-!!: Of !5; F,(^.l} fu .;uen of th; cvertemperature re;;ter tr:p.

&_.! A L ? _

'.".'he n

__A Ar _ ff _ _ __ AL_

A.. L. ._,A.r.

  • M, f ._ A.. L. ._A.._f______

......m, A.L..._m ,"r4w .....A. ...

. . . . . . . . . . . . ,,z., ,

overt;mper;tur; r;;; tor tr:p; w;lll .;du;; 'h; ;;tp;lnt; to prev ld; 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; I j

i 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 9591i confidenoe level that the hot fuel pellet in ti)e., core dosa;not experience oente.rline fueltneitingj The reactor _ core:SLs~are:used..toidefine;theivariouslRPSilunctionsisualil

' that the:above: criteria;are~. satisfied during steady; state operationgnormal operational transients; Mand anticipated; operational coeurrences'.(AOOs)g To. ensure,that.the RPS precludes theylolation;of<the'above criterial additional criteria:are applied,to the:Overtemperature N-16 reactor. trip fun.ctionsghat,lsfit must be'. demonstrated that the average enthalpy in the hot leg is less than or, equal to the saturationienthalpy_and that, tile core . exit qualityjs within;the lirnits defined by the;DNB, correlation;]

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

Reactor Core SLs B 2.1.1 BASES (continued) .

1 Appropriate; functioning of the RPS and the steem safety valves, ensure that foryariationsJn theJHERMAL _ _ ,RCS PressureRRCSl average temperaturet.RCS flow talerandMthat the reactoteprei8t.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 only MODES in which the reactor is critical. Automatic protection functions are required to be OPERABLE during MODES 1 and 2 to ensure operation within the reactor core SLs. The steam generator safety valves or automatic protection actions serve to prevent RCS heatup to the reactor core SL conditions or to initiate a reactor trip function, which forces the unit into MODE 3. Allowable Values for the reactor trip functions are specified in LCO 3.3.1, " Reactor Trip System (RTS)

Instrumentation." In MODES 3,4,5, and 6, Applicability is not required since the reactor is not generating significant THERMAL POWER.

SAFETY LIMIT The following SL violation responses are applicable to the reactor core VIOLATIONS 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 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> recognizes the importance of bringing 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 be resumed until authorized by the Commission.

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

2. FSAR, Chapter 7.

l

3.
  • Power Distribution Control Analysis and Overtemperature N-16 j and Overpower N-16 Trip Setpoint Methodology," RXE  !

006-P-A, June 1994. '

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

_ I

RTS Instrumentation 3.3.1 hatt.11 Overtemoerature N 16 The Overtemperature N 16 Function Allowable Value shall not exceed the following setpoint by more than 1.72% of span for Unit 1, or 2.82% of span for Unit 2.

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

Where:

Q,,,,,oi = Overtemperature N-16 trip setpoint, K, = 4-%O *s, K, = 9-044Bi'fF f; Uni'.1 0.01""' f:: Uni 2 K3 =

040080;*1psig-fee-Un++

Tc = Cold leg temperaturo

  • e 8 be{e,rgce T,c,_at, RATED THERMAL POWER, C'E E!E+ E Eid P = Measured pressurizer pressure, psig P' a E0863 psig (Nominal RCS operating pressure) s = the Laplace transform operator, sec-',

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% RTP Un " 0% when-662% RTP < (q,- q ) < +-6-03% RTP B-41((qi- go) 64f%)when (q,- q.) a +-6-92% RTP 1

0,404%- qd

  • SS%' " hen (;c qd SS% P "

L'n9 2: C a'  : hen 95% PT ' (;,--qJ ' 5.1 % P

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

l Note 2. Not Used.

Les specified irt the1 COLL 3 COMANCHE PEAK - UNITS 1 AND 2 3.3-21 Amendment No. 64

l RCS Pr:ssura, Temperatura, and Flow DNB Limits 3.4.1 I

i 3,4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure. Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits l

LC0 3.4.1 RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified below:

a. Pressurizer pressure 2Ethe21mitispecified';jnithelCOLR 2210 p;ig:

l

b. RCS average temperature sLthel]init;;specifjedlinithe COLR259B4: and
c. RCS total flow rate 2 403.400 3897700 gpmandi,the limitIspecifiedI1Ethe[COLRifer Unit 1 2 400,000 gp;;; for Unit 2.

APPLICABILITY: MODE 1

.......................................... NOTE -- - - -

Pressurizer pressure limit does not apply during:

a. THERMAL POWER ramp > 5% RTP per minute: or
b. THERMAL POWER step > 10% RTP.

l ACTIONS l CONDITION REQUIRED ACTION COMPLETION TIME i

l A. One or more RCS DNB A.1 Restore RCS DNB 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> l parameters not within parameter (s) to limits. within limit.

(continued) i' COMANCHE PEAK - UNITS 1 AND 2 l 3.4-1 Amendment No. 64 i

p l' RCS Pr:ssura, Temperatura, and Flow DNB Limits I- 3.4.1 l

ACTIONS (continued) .

CONDITION REQUIRED ACTION COMPLETION TIME L

B. - -

NOTE -

B.1 Maintain THERMAL Immediately POWER less than 85%

l-Only applicable prior RTP.

l to exceeding 85% RTP after a refueling outage.

Measured RCS Flow not within limits.

C. Required Action and C.1 Be in MODE 2. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1Verifypressurizerpressureis&theGimit 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> specifiedi1Mthe;COLRt 2210 p;ig.

SR 3.4.1.2VerifyRCSaveragetemperatureis$ithellfait; 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> specifjedfjnJthe;COLR2 s 502"f. i i

(continued)  !

l l^ .

I l l

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

rl SURVEILLANCE FREQUENCY SR 3.4.1.3 Verify RCS total flow rate isS&389@0;4ndh2 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> t hem i si.tispeci fi.edlj n S the'C0,LRs a 400,400 gp;r. for Un:t i

. 400,000 gpm. for Unlt 2.

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

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

18 months Verify by precision heat balance that RCS total flow rate is'k 389200 andAthejimit specified in'the_

COLR:

~ T400,400 gp,T, for Unit i

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

l 1

I l

i l

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

RCS Prassura, Tempsratura, and Flow DNB Limits B 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 safety analyses. The safety analyses (Ref.1) of normal operating conditions and anticipated operational occurrences assume initial conditions within the normal steady state envelope. The limits placed on RCS )

pressure, temperature, and flow rate ensure that the minimum departure from nucleate boiling ratio (DNBR) will be met for each of the transients analyzed.

The RCS pressure limit is consistent with operation within the nominal operational envelope. Pressurizer pressure indications are averaged to come up with a value for comparison to the limit. A lower pressure will cause the reactor core to approach DNB limits.

The RCS coolant average _ temperature limit is consistent with full power operation within the nominal operational envelope. Indications of -

temperature are averaged to determine a value for comparison to the limit. I A higher average temperature will cause the core to approach DNB limits. I i

The RCS flow rate normally remains constant during an operational fuel cycle with all pumps running. The minimum RCS flow limit corresponds to that assumed for DNB analyses and includes an allowance of 1.8% flow for measurement uncertainties. Flow rate indications from the plant computer or RCS flow rate indicators are averaged to come up with a value for comparison to the limit during shiftly surveillances. A lower RCS flow will cause the DNB limits to be approached. After each refueling, the elbow tap differential pressure transmitters are normalized to the precision RCS flow measurement. The uncertainty associated with the RCS flow measurement (1.8%)is based on the use of the feedwater venturis and precision instrumentation which has been calibrated within 90 days of performing the calorimetric flow measurement.

(continued) 4 i

4 i

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

'7 .

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

The requirements of this LCO represent the initial conditions for DNB limited transients analyzed in the plant safety analyses (Ref.1). The safety analyses have shown that transients initiated from the limits of this LCO will result in meeting the DNBR criterion. This is the acceptance limit for the RCS DNB parameters. Changes to the unit that could impact these parameters must be assessed for their impact on the DNBR criterion. The transients analyzed for include loss of coolant flow events and dropped rod events. A key assumption for the analysis of these events is that the core power distribution is within the limits of LCO 3.1.7, " Control Bank Insertion Limits"; LCO 3.2.3, " AXIAL FLUX DIFFERENCE (AFD)"; and LCO 3.2.4," QUADRANT POWER TILT RATIO (QPTR)."

The pressurizer pressure limit of 2210 p;4 and the RCS average temperature limit specifiedjn the COLR ci 502*I correspond to $a analyticallimits ef-2205 p;l;;:nd 504.7 I(for Unlt i,505.2'r for Unlt E) used in the safety analyses, with allowance for measurement uncertainty. These uncertainties are based on the use of control board indications. J 1

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 flow rate - to ensure the core operates within the limits assumed in the safety analyses. Theseyariables'are~ considered _in theLCOLR to provide operating and, analysis flexibihty,from.cycleito; cycle. . However; the minimum;RCSl flow? based on maximum 3palyzed;steegt generator tube plugging, isjetained in:the;TSACOl Operating within these limits will result in meeting the DNBR criterion in the event of a DNB limited transient.

RCS total flow rate contains a measurement error of 1.0"' based on performing a precision heat balance and using the result to normalize 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.  ;

(continued) )

i i

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

n RCS Pressure, Temperature, and Flow DNB Limits l B 3.4.1 1 BASES l

LCO Any fouling that might significantly bias the flow rate measurement can l (continued) be detected by monitoring and trending various plant performance be l 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. CPSES also uses the Transit Time Flow Meter (TTFM) to measure the volumetric RCG hot leg flow rate. The use of the TTFM results in an RCS flow measurement which is more accurate and less sensitive to RCS fluid conditions than other methods.

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

APPLICABILITY In MODE 1, the limits on pressurizer pressure, RCS coolant average temperature, and RCS flow rate must be maintained during steady state operation in order to ensure DNBR criteria will be met in the event of an unplanned loss of forced coolant flow or other DNB limited transient. In all other MODES, the power level is low enough that DNB is not a concern. j A Note has been added to indicate the limit on pressurizer pressure is not applicable during short term operational transients such as a THERMAL POWER ramp increase > 5% RTP per minute or a THERMAL POWER step increase > 10% RTP. These conditions represent short term perturbations where actions to control pressure variations might be counterproductive. Also, since they represent transients initiated from power levels < 100% RTP, an increased DNBR margin exists to offset the temporary pressure variations.

Ar,cther act of llmits or, DNS rc';ted paramcicr; The_.DNBR limit is provided in SL 2.1.1, " Reactor Core SLs." Tho;; llm ts Th.e_ conditions which define the DNBR limit are less restrictive than the limits of this LCO, b0t violatiori of 5 Safety Limit (SL) merits a stricter, more severe Required Action. Should a violation of this LCO occur, the operator must check whether or not an SL may have been exceeded.

(continued) l COMANCHE PEAK - UNITS 1 AND 2 8 3.4-3 Amendment No. 64

r

'5.6.5 CORE OPERATING LIMITS REPORT (COLR)

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:
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 for Specification 3.2.3, ,
5) Heat Flux Hot Channel Factor, K(Z), W(Z), FoRTP, and the FaC(Z) allowances for Specification 3.2.1,
6) Nuclear Enthalpy Rise Hot Channel Factor Limit and the Power j Factor 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 irl Specification _3;411
b. The analytical methods used to determine the core operating limhs shall be those previously reviewed and approved by the NRC, specifically those described 'in the following documents:
1) WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," July 1985 (W Proprietary). i
i
2) WCAP-8385," POWER DISTRIBUTION CONTROL AND LOAD l l

FOLLOWING PROCEDURES - TOPICAL REPORT," September j 1974 (E Proprietary).

L (continued)

. i L

l 4

i COMANCHE PEAK - UNITS 1 AND 2 5.0-31 Amendment No. 64 L

I R:: porting Requirements 5.6 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 Safety Analysis Aspects of an Improved Load Follow Package.

4) NUREG-0800, Standard Review Plan, U.S. Nuclear Regulatory Commission, Section 4.3, Nuclear Design, July 1981. Branch Technical Position CPB 4.3-1, Westinghouse Constant Axial Offset Control (CAOC), Rev. 2, July 1981.
5) WCAP-10216-P-A, Revision 1 A, " RELAXATION OF CONSTANT AXlAL OFFSET CONTROL Fa SURVEILLANCE TECHNICAL SPECIFICATION," February 1994 (W Proprietary).
6) WCAP-10079-P-A,"NOTRUMP, A NODAL TRANSIENT SMALL BREAK AND GENERAL NETWORK CODE," August 1985,(W Proprietary).
7) WCAP-10054-P-A, " WESTINGHOUSE SMALL BREAK ECCS EVALUATION MODEL USING THE NOTRUMP CODE", August 1985, (W Proprietary).
8) WCAP-11145-P-A, " WESTINGHOUSE SMALL BREAK LOCA ECCS EVALUATION MODEL GENERIC STUDY WITH THE NOTRUMP CODE", October 1986, (W Proprietary).
9) RXE-90-006-P, " Power Distribution Control Analysis and Overtemperature N-16 and Overpower N-16 Trip Setpoint Methodology, " February 1991,
10) RXE-88-102-P,"TUE-1 Departure from Nucleate Boiling Correlation", January 1989.
11) RXE-88-102-P, Sup.1, "TUE-1 DNB Correlation - Supplement 1",

December 1990.

(continued)

COMANCHE PEAK - UNITS 1 AND 2 5.0-32 Amendment No. 64 l

i

rn 1

,. s R: porting R:quirements 5.6 5.6 Reporting Requirements (continued) l 5.6.5 CORE OPERATING LIMITS REPORT (continued) l l

12) RXE-89-002,"VIPRE-01 Core Thermal-Hydraulic Analysis Methods for Comanche Peak Steam Electric Station Licensing Applications", June 1989.
13) RXE-91-001," Transient Analysis Methods for Comanche Peak Steam Electric Station Licensing Applications", February 1991.
14) RXE-91-002," Reactivity Anomaly Events Methodology", May 1991.

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

Methodology", December 1990.

16) TXX-88306," Steam Generator Tube Rupture Analysis", March j 15,1988.
17) RXE-91-005," Methodology for Reactor Core Response to Steamline Break Events," May,1991.
18) RXE-94-001-A," Safety Analysis of Postulated inadvertent Boron Dilution 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) 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.

(continued)

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

g ,.

Attachment 2 to TXX-99113

' Page 8 of 8 '

approved methodologies. All accident analyses, performed in accordance with these methodologies, must meet the applicable, NRC-approved limits of the safety analysis.

These changes are essentially administrative and do not change the type or quantity of effluents released offsite, nor will these changes increase individual or cumulative occupational radiation exposure. Based on the preceding evaluation, these changes do not involve a significant hazards consideration.

TXU Electric has evaluated the proposed changes and has determined that the changes do not involve (i) a significant hazards consideration, (ii) a significant change in the types or 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 exclusion set forth in 10CFR51.22(c)(9). Therefore, pursuant to 10CFR51.22(b), an environmental assessment 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's Group), dated January 19,1999 1

Lm.

Retrieved from "https://kanterella.com/w/index.php?title=ML20195B677&oldid=3042249"