Text

Revision 1 to the Cycle 20 Core Operating Limits Report
	ML16014A097
Person / Time
Site:	South Texas
Issue date:	12/29/2015
From:	Dunn R; South Texas
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NOC-AE-15003323, STI: 34256296
	Download: ML16014A097 (18)
	v • d • e

Nuclear Operating Company South Texas ProlectElectric Generating,Stati'on P.O. Box 28.9 Wadsworth, Texas 77483 ______________v_____________

December 29, 2015 NOC-AE-1 5003323 10 CFR 50.36 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 South Texas Project Unit 1 Docket No. STN 50-498 Revision to the Unit 1 Cycle 20 Core Operatinq Limits Report In accordance with Technical Specification 6.9.1 .6.d, STP Nuclear Operating Company submits the attached Core Operating Limits Report for Unit 1 Cycle 20, Revision 1. The report covers the core design changes made during the 1RE1 9 refueling outage. This revision accounts for operating with 56 versus 57 full-length rod cluster control assemblies in accordance with the license amendment No. 208 received on December 11,2015, ML15343A128.

There are no commitments in this letter.

If there are any questions regarding this report, please contact Rafael Gonzales at (361) 972-4779 or me at (361) 972-7743.

Roland F. Dunn Manager, Nuclear Fuel & Analysis rjg

Attachment:

South Texas Project Unit 1 Cycle 20 Core Operating Limits Report, Revision I

NOC-AE-1 5003323 Page 2 of 2 cc: (electronic copy)

(paper copy)

Morgqan. Lewis & Bockius LLP Regional Administrator, Region IV Steve Frantz, Esquire U.S. Nuclear Regulatory Commission 1600 East Lamar Boulevard U.S. Nuclear Regulatory Commission Arlington, TX 76011-4511 Lisa M. Regner Lisa M. Regner NRG South Texas LP Senior Project Manager John Ragan U.S. Nuclear Regulatory Commission Chris O'Hara One White Flint North (O8H04) Jim von Suskil 11555 Rockville Pike Rockville, MD 20852 CPS Enerqy Kevin Polio NRC Resident Inspector Cris Eugster U. S. Nuclear Regulatory Commission L. D. Blaylock P. O. Box 289, Mail Code: MNl16 Wadsworth, TX 77483 Cramn Caton & James, P.C.

Peter Nemeth City of Austin Cheryl Mele John Wester Texas Dept. of State Health Services Richard A. Ratliff Robert Free

Nuclear Operating Company SOUTH TEXAS PROJECT Unit 1 Cycle 20 CORE OPERATING LIMITS REPORT Revision I Core 0Operafing Limits ReportPge1oi Page i of 16

N~tlBl~ *--Unit l*_~l"-" 1 Cycle 20 Nuclear operating Company Core Operating Limits Report Rev. 1

l* ,.Page~ ofl16 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report for STPEGS Unit 1 Cycle 20 has been prepared in accordance with the requirements of Technical Specification 6.9.1.6. The core operating limits have been developed using the NRC-approved methOdologies specified in Technical Specification 6.9.1.6.

Thle Technical Specifications affected by this report ar'e:

1) 2.1 SAFETY LIMITS

2) 2.2 LIMITING SAFETY SYSTEM SETTINGS

3) 3/4.1.1.1 SHUTDOWN MARGIN

4) 3/4.1. 1.3 MODERATOR TEMPERATURE COEFFICIENT LIMITS

5) 314.1i.3.5 SHUTDOWN ROD INSERTION LIMITS

6) 3/4.1.3.6 CONTROL ROD INSERTION LIMITS

7) 314.2.1 AED LIMITS

8) 3/.4.2.2 HEAT FLUX HOT CHANNEL FACTOR

9) 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR

10) 3/4.2.5 DNB PARAMETERS 2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are pt-esented below.

2.1 SAFETY LIMITS (Specification 2.1.):

2.1. 1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tavg) shall not exceed the limits shown in Figure 1.

2.2 LIMITING SAFETY SYSTEM SETTINGS (Specification 2.2):

2.2.1 221 The Loop design flow for Reactor Coolant Flow-Low is 98,000 gpm.

NucerOperatng Company Ui C Opyle 20inLimilts Report Rev. 1 2.2.2 The Over-temperature AT and Over-power AT setpoint parameter values are listed below:

Over-temperature AT Setpoint Parameter Values tim measured reactor vessel AT lead/lag time constant, ti 8 sec "T2 measured reactor vessel AT lead/lag time constant, t'2 =3 sec

-t3 measured reactor vessel AT lag time constant, "c3: 2 see "c4 measured reactor vessel average temperature lead/lag time donstant, t4 =28 sec "t5 measured reactor vessel average temperature lead/lag time constant, '5 =4 sec "t6 measured reactor vessel average temperature lag time constant, "¶6= 2 sec K1 Overtemperature AT reactor trip setpoint, K1 = 1.14 K(2 Overtemperature AT reactor trip setpoint Taivg coefficient, K1(2 0.028/°F K3 Overtemperature AT reactor trip setpoint pressure coefficient, K3 0.00143/psi T' Nominal full power T*vg T'_* 592.0 0 F P' Nominal RCS pressure, P' = 2235 psig fi(AI) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(1) For qL - qb between -70% and +8%, f, (Al) =0, where qt and qb are percent RATED THERMAL POWER in thetop and bottom halves of the core respectively, and qt + qu is total THERMAL POWER in percent of RATED THERMAL POWER; (2) For each percent that the magnitude of qt - qu exceeds -70%, the AT Trip Setpoint shall be automatically reduced by 0.0% of its value at RATED THERMAL POWER; and (3) For each percent that the magnitude of qt - qb exceeds +8%, the AT Trip Setpoint shall be automatically reduced by 2.65% of its value at RATED THERMAL POWER. (Reference 3.6 and Section 4.4.1.2 of Reference 3.7)

Over-power AT Setpoint Parameter Values tj measured reactor vessel AT leadilag time constant, 'ri = 8 sec

"¶2 measured reactor vessel AT leadllag time constant, ¶2 =3 sec

"¶3 measured reactor vessel AT lag time constant, ¶*3 =2 sec

"¶6 measured reactor vessel average temperature lag time constant, ¶6= 2 sec

-ri Time constant utilized in the rate-lag compensator for Tavg, = 10 sec

¶-7 K4 Overpower AT reactor trip setpoint, K4 1.08 Ki Overpower AT reactor trip setpoint T.avg rate/lag coefficient, Ks = 0.02/0 F for" increasing average temperature, and K5 = 0 for decreasing average temperature K6 Overpower AT reactor trip setpoint Tavg heatup coefficient Ks = 0.002/°F for T>T".,andKQ OforT_< T" T" Indicated full power Tav*, T"_< 592.0 0 F f,_(AI) =0 for all (Al)

Nucl lT Operating M

earCompany Unit 1Cycle 20 Cor'e Operating Limits Report Rev. 1 I Jr"age4 of 16 2.3 ShtUTDOWN MARGIN' (Specification 3.1.1,1):

The SHUTDOWN MARGTN shall be:

2.3.1 Greater than 1.3% Ap for MODES 1 and 2*

See Special Test Exception 3.10.1 2.3.2 Greater than the limits in Figure 2 for MODES 3 and 4.

2.3.3 Greater than the limits in Figure 3 for MODE 5.

2.4 MODERATOR TEMPERATURE COEFFICIENT (Specification 3.1.1.3):

2.4.1 The BOL, ARO, MTC shall be less positive than the limits shown in Figure 4.

2.4.2 The EOL, ARO, HFP, MTC shall be less negative than -62.6 pcmr/°F.

2.4.3 The 300 ppm, ARO, HEP, MTC shall be less negative than -53.6 pcrn/°F (300 ppm Surveillance Limit).

Where: BOL stands for Beginning-of-Cycle Life, EOL stands for End-of-Cycle Life, ARO stands for All Rods Out, HIT stands for Hot Full Power (100% RATED THERMAL POWER),

I-FP vessel average temperature is 592 °F.

2.4.4 The Revised Predicted near-EOL 300 ppm MTC shall be calculated using the algorithm from the document referenced by Technical Specification 6.9.1 .6.b.lI0:

Revised Predicted MTC =Predicted MTC + AFD Correction - 3 pcmi°F If the Revised Predicted MTC is less negative than the COLR Section .2.4.3 limit and all of the benchmuark data contained in the surveillance procedure are met, then an MTC measurement in accordance with S.R. 4.1.1 .3b is not required.

2.5 ROD INSERTION LIMITS' (Specification 3.1.3.5 and 3.1.3.6):

2.5.1 All banks shall have the same Full Out Position (.FOP) of either 254 or 259 steps withdrawn.

2.5.2 The Control Banks shall be limited in physical insertion as specified in Figure 5.

2.5.3 Individual Shutdown bank rods are fully withdr-awn when the Bank Demand Indication is at the FOP and the Rod Group: Height Limiting Condition for Operation is satisfied (T:S. 3.1.3.1).

I The Shutdown Margin and Rod Insertion limits account for the removal of RCCA D6 in Shutdown Bank A.

Opert n Cm-"y .'omp..y Core5 Limits Report CUnitOperating RV r 2.6 AXIAL FLUX DIIFFERENCE (Specification 3.2.1):

2.6.1 AFD limits as required by Technical Specification .3.2.1i are, determined by Constant Axial Offset Control (CAOC) Operations with an AFD target band of +5, -10%.

2.6.2 The AFD shall be maintained within the ACCEPTABLE OPERATION portion of Figure 6, as required by Technical Specifications.

2.7 HEAT FLUX HOT CHANNEL FACTOR (Specification 3.2.2):

2.7.1 F*TP 2.55.

2.7.2 K(Z) is provided in Figure 7.

2,7.3 The Fx limits for RATED THERMAL POWER (pFPR")Within specific core planes shall be:

2.7.3o1 Less than or equal to 2.102 for all cYCle burnups for all core :planes containing Bank "D"' control rods, and 2.7:.3.2 Less than or' equal to the appropriate core height-dependent value from Table 1 for all unrodded core planes.

2.7.3.3 PFxy = 0.2, These FPy limits were.used to confirm that the heat flux hot channel factor F0 (Z) will be limited by Technical Specification 3.2.2 assuming the most-limiting axial power distributions expected to result for the insertion and removal of Control Banks C and D during operation, including the accompanying variations in the axial xenon and power distributions, as described in WCAP-83 85. Therefore, these F.y limits provide assurance that the initial conditions assumed in the LOCA analysis are met, alo~ng with thle ECCS acceptance criteria of!10 CFR 50.46.

2.7.4 Core Power Distribution Measurement Uncertainty for the Heat Flux Hot Channel Factor 2.7.4.1 If the Power Bistribultion Monitoring System (PDMS) is operable, as defined in the Technical Requirements Manual Section 3.3.3.12, the core power distrnbution measurement uncertainty (UFQ) to be applied to the FQ(Z) and F.*y(Z) using the PDMS shall be calculated by:

UFQ = (1.0 + (UQ/100))*UE Where:

UQ = Uncertainty forpower peaking factor as defined in Equation 5-19 fron the document referenced by Technical Specification 6.9.t.6.b.11I UE = Engineering uncertainty factor of 1.03.

This uncertainty is calculated and applied automatically by the Power Distribution Monitoring System (PDMS),

UnitleaCrle 20 LmitsnReprt Rcvy mr * ~~~~~Core Operating iisRproe, Page 6of 1.6 2;7.4.2 If the moveable detector system is used, the core power distribution measurement uncertainty (UFQ) to be applied to the FQ*(Z) and Fxy(Z) shall be calculated by:

UFQ = UQU*TJUE Where:

UQU =Base EQ measurement uncertainty of 1.05.

UE = Engineering uncertainty factor of 1.03=.

2.8 ENTHALPY RISE HOT CHANNEL FACTOR (~Specification 3.2.3):

2,8.1 F§i'= 1.62 2.8.2 PFAxH =0.3 2.8.3 Core Power Distribution Measuremcnt Uncertainty for the Enthalpy Rise H-ot Channel Factor 2.8.3.1 If the Power Distribution Monitoring System (PDMS) is operable, as defined in the Technical Requirements Manual Section 3...2 tecr power distribution me~asurement uncertainty (UiixH) to be aplied to the~ Fh using the PDMS shall be the greater of:

UFAII = 1.104 OR UF*-I =1.0+ (UA,/100)

Where:

UAH = Uncertainty for power peaking factor as defined in Equation 5-19 from' the document referenced in Technical Specification 6.9.1 .6.b.11!.

This uncertainty is calculated and applied automatically by the Power Distribution Monitoring System.

2.8.3.2 If thae moveable detector system is used, the core power distribution measurement uncertainty (UFAH) shall be:

UFAH --1.04

lll~

l*Unit I Cycle 20 Nu1ea Opratn Crpn Core Operating Limits Report Rev. 1 page 7 of I6 2.9 DNB PARAMETERS (Specification 3.2.5):

2.9.1 The following DNB-related parameters shall be maintained within the following limits:2 2.9.1.1 Reactor Coolant System Tlavg _<595 0 F 3, 2.9.1.2 Pressurizer Pressure > 2200 psig",

2..9.1.3 Minimum Measured Reactor Coolant System Flow > 403,000 gpm5.

3.0 REFERENCES

3.1 Letter from J. M. Ralston (Westinghouse) to D. F. Hoppes (STPNOC), "South Texas Project Electric Generating Station Unit I Cycle 20 Redesigna Final Reload Evaluation" NF-TG-15-62, Rev. I (ST-UB-NOC-1 5003490, Rev. 1) dated November 25, 2015.

3.2 NUREG-1346, Technical Specifications, South Texas Project Unit Nos. 1 and 2.

3.3 STPNOC Calculation ZC-7035, Rev. 2, "'Loop Uncertainty Calculation for RCS Tavg Instrumentation," Section 10.1.

3.4 STPNOC Calculation ZC-7032, Rev. 6, "Loop Uncertainty Calculation for Narrow Range Pressurizer Pressure Monitoring Instrumentation," Section 2.3, Page 9.

3.5 5Z529ZB01025.Rev. 4, Design Basis Document, Technical Specifications /LCO, Tech Spec Section 3.2.5.c.

3.6 Letter from J. M. Ralston (Westinghouse) to D. F. Hoppes (STPNOC), "South Texas Project Electric Generating Station Units 1 and 2 DocUmentation of the ft (Al) Function in OTAT Setpoint Calculation," NP-TG-I11-93 (ST-UB-NOC-l 100321[5) dated November 10, 20l1, 3.7 Document RSE-U1, Rev. 5, "Unit 1 Cycle 20 Reload Safety Evaluation~ and Core Operating Limits Report." (CR Action 14-1i0332-67) 2A discussion of the processes to be used to take these readings is provided in the basis for Technfical Specification 3.2.5.

SIncludes a !.9 0 F measurement tmcertainty per Reference 3.3, Page 37.

'~Limit not applicable during either a Thermal Power ramp in.excess of 5%of RTP per nminute or a Thermal POwer step in excess of 10% RTP. Per Technical Specification 3.2.5 Bases, this includes a 10.7 psi measurement uncertainty as read on thle QDPS display, which is bounded by the 9.6 psi averaged measurement calculated in Reference 3A.

Inhcludes the most limiting flow measurement uncertainty of 2.8% from Reference 3.5.

Unit 1 Cycle 20 Nuclear Operating Company Core Operating Limits Report Rev. 1 Page 8 of 16 Figure 1 Reactor Core Safety Limits - Four Loops in Operation 680 a

c,2 0 20 40 60 80 100 120 140 Rated Thermal Power (%)

N~OingC*in Unit 1 Cycle 20 Lmt eotRv Nucler CopanyCore Opratig Operating Lmt eotRv m r Page 9 of 16 Figure 2 Required Shutdown Margin for Modes 3 & 4 7.0 6.0 5.0

.. *4.0

~3.0 I-2.0 1.0 0.0 0 400 800 1200 1600 2000 2400 RCS Critical Boron Concentration (ppm)

(for ARE minus most reactive stuck rod)

diT Mr Nuclea Oprtn Copn Unt Cycle 20 Core Operating Limits Report Rev. I Pagl10 of 16 Figure 3 Required Shutdown Margin for Mode 5 7.0 6.0 5.0

=*4.o

= 3.0 2.0 1.0 0.0 0400 800 1200 1600 2000 2400 RCS Critical Boron Concentration (ppm)

(for ART minus most reactive stuck rod)

,,4 !,roper!!g'Compny UCote1Operleing Limnits Report Rev. 1

1*

r Page t11of" 6 Figure 4 MTC versus Power Level 7.0 6.0 Unacceptable 5.0 p"4.0 Acceptable1 I,d S2.0

~0.0

-1.0

-2.0

-3.0 0 10 20 30 40 50 60 70 80 90 100 Rated Thermal.Powser (%)

'* 1*Unit I Cycle 20 Nuclear operating Company Core Operating Limits Report Rev. 1 Pagc 12 of 16 Figure 5 Control Rod Insertion Limits* versus Power Level 260 llll)l[l*ltk*l lll[ll[llll)llJ*

(i2"3,-2-59):- 122-Step ...

Overlap I-4 122 Step Overlap H(21,254): 117 Step Overlap1 ] -A(77,254 )* 117 Step Overlap

_(79,259);

I 1 I II I .A'l I * , .1* . ., *7..... ; t I I I I I I 1J I I ,-

240 I 220 200 180 160 r* 140 (,,65 S120 Coto~akAisarad ihdantoulOu o itin S100 80 60 40 20 0

0 10 20 30 40 50 60 70 80 90 100 Rated Thermal Power (%)

Nuclear Operating Company UniteIOCyclei20 Limits Report Rex'. 1 Coe peatngPage 13of1 Figure 6 AFD Limits versus Power Level 120 110 , , , : I 1_ .

100 2K- -- 9oI 90 80

... .. , I _ _ I,9o 70 I.7- . . . .

I-60 50

(-31

,5 ) -*-,-4 31 ,50) 40 30 20 10 0

-50 -40 -30 -20 -10 0 10 20 30 40 50 Axial Flux Difference (% Delta-I)

SlmT~iUnit ICycle 20 Nuclear Operating Company Core Operating Limits Report Rev. 1 Page 14 of 16 Figure 7 K(Z) - Normalized FQ(Z) versus Core Height 1.2 .. .2 i............ I I'5

t i v F,*. ... IW

I ........ : IIi T T_.. ..........

!.i .. . .I ,.... .. ........

1.0 0.9 ilijiiiiiiV~~ii~i~l T'..!..L.!... 2

_ _ , _*_= ,

0.8 ..............

...... .................... * ,I l ! i i l -

-ii*oele~t -T FQI* K(Z)-.-F~i i

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..................... 1, 2. II i i . ........... ............ ..IJ.. ..

0...............1 ........ 2...... ...... 314 .0 7.5 0 .9 2 10 I]:I!;11 12....... 13................ 14}:]i:i 0.2 ...... .

.......................... re Heig ht................ ' [Co .

0.1 0.0

e lalpUnit I Cycle 20 Nuclear Operatlng Company Core Operating Limits Report Rev. 1 I rPage 15 of16 Table 1 (Part 1 of 2)

Unrodded Fx.y for Each Core Height for Cycle Burnuips Less Than 9000 M'WDIMTU Core Height Axial Unrodded Core Height Axial Unrodded

(.Ft.) Point Fxy (Ft.) Point Fxy 14.0 1 7.3 17 6.8 37 1.972 13.8 2 5.739 6.6 38 1.993 13.6 3 4.158 6.4 39 1.975 13.4 4 2.789 6.2 40 1.932 13.2. 5 2.495 6.0 41 1.901 13.0 6 2.226 5.8 42 1.938 12.8 7 2.138 5.6 43 1.952 12.6 8 2.118 5.4 44 1.958 12.4 9 2.068 5.2 45 2.001 12.2 10 2.022 5.0 46 2.058 12.0 11 2.0 02 4.8 47 2.063 11.8 12 2.014 4.6 48 2.009 11.6 13 2.037 4.4 49 1.946 11.4 14 2.014 4.2 50 1.966 11.2 15 1.965 4.0 51 1.973 11.0 16 1.933 3.8 52 1.965 10.8 17 1.925 3.6 53 1.974 10.6 18 1.920 3.4 54 2.014 10.4 19 1.918 3.2 55 2.039 10.2 20 1.938 3.0 56 1.991 10.0 21 1.973 2.8 57 1.93 8 9.8 22 1.981 2.6 58 1.942 9.6 23 i1.943 2.4 59 1.947 9.4 24 1.908 2.2 60 1.952 9.2 25 1.904 2.0 61 1.967 9.0 26 1.895 1.8 62 1.997 8.8 27 1.896 1.6 63 2.004 8.6 28. 1.9 16 1.4 64 1.932 8.4 29 1.983 1.2 65 1.872 8.2 30 2.036 1.0 66 1.912 8.0 31 1.977 0.8 67 2.222 7.8 32 1.925 0.6 68 3.005 7.6 33 1.929 0.4 69 4.318 7.4 34 1.945 0.2 70 6.145 7.2 35 1.947 0.0 71 9.180 7.0 36 1.939

N*rOPll* n Unit I Cycle 20 Lmt eotRv Nuclar peraingC~omanyCore Operating Lmt eotRv Page 16 of 16 Table 1 (Part 2 of 2)

Unrodded Fxy for Each Core Height for Cycle Burntips Greater Than or Equal to 9000 MWD!MTU Core Height Axial Unrodded Core Height Axial Unrodded (Ft.)_ Point Fxy (Ft.) Point Fxy 14.0 1 6.495 6.8, 37 2.203 13.8 5.203 6.6 38 2.238 13.6 3 3.911 6.4 39 2.204 13.4 4 2.790 6.2 40 2.147 13.2 5 2.552 6.0 41 2.114 13.0 6 2.305 5.8 42 2,106 12.8 7 2.160 5.6 43 2.095 12.6 8 2.087 5.4 44 2.08 1 12.4 9 2.027 5.2 45 2.098 12.2 t0 2.021 5.0 46 2.131 12.0 11 *2.030 4.8 47 2.125 11.8 12 2.056 4.6 48 2.067 11.6 13 2.082 4.4 49 2.02 1 11.4 14 2.074 4.2 50 2,016 11.2 15 2,044 4.0 51 2.005 11.0 16 2.009 3.8 52 1.990 10.8 17 2.038 3.6 53 1,992 10,6 18 2.046 3.4 54 2.026 10.4 19 2.053 3.2 55 2.048 10.2 20 2.082 3.0 56 1.990 10.0 21 2.127 2.8 57 1,935 9.8 22 2.145 2.6 58 1.905 9.6 23 2.113 2.4 59 1.875 9.4 24 2.088 2.2 60 1.864 9.2 25 2.102 2.0 61 1.880 9.0 26 2.110 1.8 62 1.926 8.8 27 2,114 1.6 63 1.954 8.6 28 2.124 1.4 64 1.932 8.4 29 2.161 1.2 65 1.945 8.2 30 2.194 1.0 66 2.054 8.0 31 2.157 0.8 67 2.4 18 7.8 32 2.124 0.6 68 3.143 7.6 33 2.126 0.4 69 4.250 7.4 34 2.137 0.2 70 5.782 7.2 35 2.148 0.0 71 8.469 7.0 36 2.160

Nuclear Operating Company South Texas ProlectElectric Generating,Stati'on P.O. Box 28.9 Wadsworth, Texas 77483 ______________v_____________

December 29, 2015 NOC-AE-1 5003323 10 CFR 50.36 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 South Texas Project Unit 1 Docket No. STN 50-498 Revision to the Unit 1 Cycle 20 Core Operatinq Limits Report In accordance with Technical Specification 6.9.1 .6.d, STP Nuclear Operating Company submits the attached Core Operating Limits Report for Unit 1 Cycle 20, Revision 1. The report covers the core design changes made during the 1RE1 9 refueling outage. This revision accounts for operating with 56 versus 57 full-length rod cluster control assemblies in accordance with the license amendment No. 208 received on December 11,2015, ML15343A128.

There are no commitments in this letter.

If there are any questions regarding this report, please contact Rafael Gonzales at (361) 972-4779 or me at (361) 972-7743.

Roland F. Dunn Manager, Nuclear Fuel & Analysis rjg

Attachment:

South Texas Project Unit 1 Cycle 20 Core Operating Limits Report, Revision I

NOC-AE-1 5003323 Page 2 of 2 cc: (electronic copy)

(paper copy)

Morgqan. Lewis & Bockius LLP Regional Administrator, Region IV Steve Frantz, Esquire U.S. Nuclear Regulatory Commission 1600 East Lamar Boulevard U.S. Nuclear Regulatory Commission Arlington, TX 76011-4511 Lisa M. Regner Lisa M. Regner NRG South Texas LP Senior Project Manager John Ragan U.S. Nuclear Regulatory Commission Chris O'Hara One White Flint North (O8H04) Jim von Suskil 11555 Rockville Pike Rockville, MD 20852 CPS Enerqy Kevin Polio NRC Resident Inspector Cris Eugster U. S. Nuclear Regulatory Commission L. D. Blaylock P. O. Box 289, Mail Code: MNl16 Wadsworth, TX 77483 Cramn Caton & James, P.C.

Peter Nemeth City of Austin Cheryl Mele John Wester Texas Dept. of State Health Services Richard A. Ratliff Robert Free

Nuclear Operating Company SOUTH TEXAS PROJECT Unit 1 Cycle 20 CORE OPERATING LIMITS REPORT Revision I Core 0Operafing Limits ReportPge1oi Page i of 16

N~tlBl~ *--Unit l*_~l"-" 1 Cycle 20 Nuclear operating Company Core Operating Limits Report Rev. 1

l* ,.Page~ ofl16 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report for STPEGS Unit 1 Cycle 20 has been prepared in accordance with the requirements of Technical Specification 6.9.1.6. The core operating limits have been developed using the NRC-approved methOdologies specified in Technical Specification 6.9.1.6.