NOC-AE-15003323, Revision 1 to the Cycle 20 Core Operating Limits Report: Difference between revisions
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Unit 1 Cycle 20 Lmt eotRv Nucler Opratig CopanyCore Operating Lmt eotRv m r Page 9 of 16 Figure 2 Required Shutdown Margin for Modes 3 & 4 7.0 6.0 5.0..~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 Unt Cycle 20 Nuclea Oprtn Copn Core Operating Limits Report Rev. I Pagl 10 of 16 Figure 3 Required Shutdown Margin for Mode 5 7.0 6.0 5.0= 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) | Unit 1 Cycle 20 Lmt eotRv Nucler Opratig CopanyCore Operating Lmt eotRv m r Page 9 of 16 Figure 2 Required Shutdown Margin for Modes 3 & 4 7.0 6.0 5.0..~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 Unt Cycle 20 Nuclea Oprtn Copn Core Operating Limits Report Rev. I Pagl 10 of 16 Figure 3 Required Shutdown Margin for Mode 5 7.0 6.0 5.0= 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. 1r Page t11 of" 6 Figure 4 MTC versus Power Level 7.0 6.0 5.0 p"4.0 I,d S2.0~0.0 Unacceptable Acceptable1 | ,, 4!,roper!!g'Compny UCote1Operleing Limnits Report Rev. 1r Page t11 of" 6 Figure 4 MTC versus Power Level 7.0 6.0 5.0 p"4.0 I,d S2.0~0.0 Unacceptable Acceptable1 | ||
-1.0-2.0-3.0 0 10 20 30 40 50 60 70 80 90 100 Rated Thermal.Powser | -1.0-2.0-3.0 0 10 20 30 40 50 60 70 80 90 100 Rated Thermal.Powser | ||
(%) | (%) | ||
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 (i2"3,-2-59):- | 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 (i2"3,-2-59):- | ||
122-Step ...Overlap H(21,254): | 122-Step ...Overlap H(21,254): | ||
117 Step Overlap 1] | 117 Step Overlap 1] | ||
I-4 _(79,259);-A( 77,254 122 Step Overlap 117 Step Overlap I 1 I I I I .A 'l I , .., ; t I I I I I I 1J I I ,-240 220 200 180 160140 S120 S100 80 60 40 20 0 I (,,65 Coto~akAisarad ihdantoulOu o itin 0 10 20 30 40 50 60 Rated Thermal Power (%)70 80 90 100 Nuclear Operating Company UniteIOCyclei20 Limits Report Rex'. 1 Coe peatngPage 1 3of1 Figure 6 AFD Limits versus Power Level 120 110 100 90 80 I-70 60 50 40 30 20 10 0 , , , : I 1_ .2K ---9oI... .. , I _ _ I,9o I.7- ....(-31 ,5 ) 31 ,50)-50-40 20 -10 0 10 20 30 40 50 Axial Flux Difference | I-4 _(79,259);-A( 77,254 122 Step Overlap 117 Step Overlap I 1 I I I I .A 'l I , .., ; t I I I I I I 1J I I ,-240 220 200 180 160140 S120 S100 80 60 40 20 0 I (,,65 Coto~akAisarad ihdantoulOu o itin 0 10 20 30 40 50 60 Rated Thermal Power (%)70 80 90 100 Nuclear Operating Company UniteIOCyclei20 Limits Report Rex'. 1 Coe peatngPage 1 3of1 Figure 6 AFD Limits versus Power Level 120 110 100 90 80 I-70 60 50 40 30 20 10 0 , , , : I 1_ .2K ---9oI... .. , I _ _ I,9o I.7- ....(-31 ,5 ) 31 ,50)-50-40 20 -10 0 10 20 30 40 50 Axial Flux Difference | ||
(% Delta-I) | (% Delta-I) | ||
SlmT ~iUnit I Cycle 20 Nuclear Operating Company Core Operating Limits Report Rev. 1 Page 14 of 16 1.2!.i 1.0 0.9 0.8 0.0.6~0.4 0.3 0.2 0.1 0.0 Figure 7 K(Z) -Normalized FQ(Z) versus Core Height I .. I'5 .2 i............ | SlmT ~iUnit I Cycle 20 Nuclear Operating Company Core Operating Limits Report Rev. 1 Page 14 of 16 1.2!.i 1.0 0.9 0.8 0.0.6~0.4 0.3 0.2 0.1 0.0 Figure 7 K(Z) -Normalized FQ(Z) versus Core Height I .. I'5 .2 i............ | ||
| Line 116: | Line 116: | ||
Unit 1 Cycle 20 Lmt eotRv Nucler Opratig CopanyCore Operating Lmt eotRv m r Page 9 of 16 Figure 2 Required Shutdown Margin for Modes 3 & 4 7.0 6.0 5.0..~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 Unt Cycle 20 Nuclea Oprtn Copn Core Operating Limits Report Rev. I Pagl 10 of 16 Figure 3 Required Shutdown Margin for Mode 5 7.0 6.0 5.0= 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) | Unit 1 Cycle 20 Lmt eotRv Nucler Opratig CopanyCore Operating Lmt eotRv m r Page 9 of 16 Figure 2 Required Shutdown Margin for Modes 3 & 4 7.0 6.0 5.0..~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 Unt Cycle 20 Nuclea Oprtn Copn Core Operating Limits Report Rev. I Pagl 10 of 16 Figure 3 Required Shutdown Margin for Mode 5 7.0 6.0 5.0= 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. 1r Page t11 of" 6 Figure 4 MTC versus Power Level 7.0 6.0 5.0 p"4.0 I,d S2.0~0.0 Unacceptable Acceptable1 | ,, 4!,roper!!g'Compny UCote1Operleing Limnits Report Rev. 1r Page t11 of" 6 Figure 4 MTC versus Power Level 7.0 6.0 5.0 p"4.0 I,d S2.0~0.0 Unacceptable Acceptable1 | ||
-1.0-2.0-3.0 0 10 20 30 40 50 60 70 80 90 100 Rated Thermal.Powser | -1.0-2.0-3.0 0 10 20 30 40 50 60 70 80 90 100 Rated Thermal.Powser | ||
(%) | (%) | ||
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 (i2"3,-2-59):- | 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 (i2"3,-2-59):- | ||
122-Step ...Overlap H(21,254): | 122-Step ...Overlap H(21,254): | ||
117 Step Overlap 1] | 117 Step Overlap 1] | ||
I-4 _(79,259);-A( 77,254 122 Step Overlap 117 Step Overlap I 1 I I I I .A 'l I , .., ; t I I I I I I 1J I I ,-240 220 200 180 160140 S120 S100 80 60 40 20 0 I (,,65 Coto~akAisarad ihdantoulOu o itin 0 10 20 30 40 50 60 Rated Thermal Power (%)70 80 90 100 Nuclear Operating Company UniteIOCyclei20 Limits Report Rex'. 1 Coe peatngPage 1 3of1 Figure 6 AFD Limits versus Power Level 120 110 100 90 80 I-70 60 50 40 30 20 10 0 , , , : I 1_ .2K ---9oI... .. , I _ _ I,9o I.7- ....(-31 ,5 ) 31 ,50)-50-40 20 -10 0 10 20 30 40 50 Axial Flux Difference | I-4 _(79,259);-A( 77,254 122 Step Overlap 117 Step Overlap I 1 I I I I .A 'l I , .., ; t I I I I I I 1J I I ,-240 220 200 180 160140 S120 S100 80 60 40 20 0 I (,,65 Coto~akAisarad ihdantoulOu o itin 0 10 20 30 40 50 60 Rated Thermal Power (%)70 80 90 100 Nuclear Operating Company UniteIOCyclei20 Limits Report Rex'. 1 Coe peatngPage 1 3of1 Figure 6 AFD Limits versus Power Level 120 110 100 90 80 I-70 60 50 40 30 20 10 0 , , , : I 1_ .2K ---9oI... .. , I _ _ I,9o I.7- ....(-31 ,5 ) 31 ,50)-50-40 20 -10 0 10 20 30 40 50 Axial Flux Difference | ||
(% Delta-I) | (% Delta-I) | ||
SlmT ~iUnit I Cycle 20 Nuclear Operating Company Core Operating Limits Report Rev. 1 Page 14 of 16 1.2!.i 1.0 0.9 0.8 0.0.6~0.4 0.3 0.2 0.1 0.0 Figure 7 K(Z) -Normalized FQ(Z) versus Core Height I .. I'5 .2 i............ | SlmT ~iUnit I Cycle 20 Nuclear Operating Company Core Operating Limits Report Rev. 1 Page 14 of 16 1.2!.i 1.0 0.9 0.8 0.0.6~0.4 0.3 0.2 0.1 0.0 Figure 7 K(Z) -Normalized FQ(Z) versus Core Height I .. I'5 .2 i............ | ||
Revision as of 12:18, 27 April 2019
| ML16014A097 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 12/29/2015 |
| From: | Dunn R F South Texas |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NOC-AE-15003323, STI: 34256296 | |
| Download: ML16014A097 (18) | |
Text
Nuclear Operating Company South Texas Prolect Electric 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 1 RE1 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: (paper copy)Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 1600 East Lamar Boulevard Arlington, TX 76011-4511 Lisa M. Regner Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North (O8H04)11555 Rockville Pike Rockville, MD 20852 NRC Resident Inspector U. S. Nuclear Regulatory Commission P. O. Box 289, Mail Code: MNl16 Wadsworth, TX 77483 (electronic copy)Morgqan. Lewis & Bockius LLP Steve Frantz, Esquire U.S. Nuclear Regulatory Commission Lisa M. Regner NRG South Texas LP John Ragan Chris O'Hara Jim von Suskil CPS Enerqy Kevin Polio Cris Eugster L. D. Blaylock 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~ 1 Cycle 20 Nuclear operating Company Core Operating Limits Report Rev. 1,.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 yle Op 20in Limilts 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 K 1 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'_ 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, =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, ¶-7 = 1 0 sec 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 K 6 Overpower AT reactor trip setpoint Tavg heatup coefficient Ks = 0.002/°F for T>T".,andKQ O forT_< T" T" Indicated full power T"_< 592.0 0 F f,_(AI) =0 for all (Al) lT M Unit 1 Cycle 20 Nucl Operating Company 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.
.'omp..y CUnitOperating Limits Report RV Opert n Cm-"y Core5 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 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 F 0 (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 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 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= 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~ 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 3 A.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 (%)
Unit 1 Cycle 20 Lmt eotRv Nucler Opratig CopanyCore Operating Lmt eotRv m r Page 9 of 16 Figure 2 Required Shutdown Margin for Modes 3 & 4 7.0 6.0 5.0..~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 Unt Cycle 20 Nuclea Oprtn Copn Core Operating Limits Report Rev. I Pagl 10 of 16 Figure 3 Required Shutdown Margin for Mode 5 7.0 6.0 5.0= 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. 1r Page t11 of" 6 Figure 4 MTC versus Power Level 7.0 6.0 5.0 p"4.0 I,d S2.0~0.0 Unacceptable Acceptable1
-1.0-2.0-3.0 0 10 20 30 40 50 60 70 80 90 100 Rated Thermal.Powser
(%)
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 (i2"3,-2-59):-
122-Step ...Overlap H(21,254):
117 Step Overlap 1]
I-4 _(79,259);-A( 77,254 122 Step Overlap 117 Step Overlap I 1 I I I I .A 'l I , .., ; t I I I I I I 1J I I ,-240 220 200 180 160140 S120 S100 80 60 40 20 0 I (,,65 Coto~akAisarad ihdantoulOu o itin 0 10 20 30 40 50 60 Rated Thermal Power (%)70 80 90 100 Nuclear Operating Company UniteIOCyclei20 Limits Report Rex'. 1 Coe peatngPage 1 3of1 Figure 6 AFD Limits versus Power Level 120 110 100 90 80 I-70 60 50 40 30 20 10 0 , , , : I 1_ .2K ---9oI... .. , I _ _ I,9o I.7- ....(-31 ,5 ) 31 ,50)-50-40 20 -10 0 10 20 30 40 50 Axial Flux Difference
(% Delta-I)
SlmT ~iUnit I Cycle 20 Nuclear Operating Company Core Operating Limits Report Rev. 1 Page 14 of 16 1.2!.i 1.0 0.9 0.8 0.0.6~0.4 0.3 0.2 0.1 0.0 Figure 7 K(Z) -Normalized FQ(Z) versus Core Height I .. I'5 .2 i............
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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 (.Ft.)Axial Point Unrodded Fxy Core Height (Ft.)Axial Point Unrodded Fxy 14.0 13.8 13.6 13.4 13.2.13.0 12.8 12.6 12.4 12.2 12.0 11.8 11.6 11.4 11.2 11.0 10.8 10.6 10.4 10.2 10.0 9.8 9.6 9.4 9.2 9.0 8.8 8.6 8.4 8.2 8.0 7.8 7.6 7.4 7.2 7.0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28.29 30 31 32 33 34 35 36 7.3 17 5.739 4.158 2.789 2.495 2.226 2.138 2.118 2.068 2.022 2.0 02 2.014 2.037 2.014 1.965 1.933 1.925 1.920 1.918 1.938 1.973 1.981 i1.943 1.908 1.904 1.895 1.896 1.9 16 1.983 2.036 1.977 1.925 1.929 1.945 1.947 1.939 6.8 6.6 6.4 6.2 6.0 5.8 5.6 5.4 5.2 5.0 4.8 4.6 4.4 4.2 4.0 3.8 3.6 3.4 3.2 3.0 2.8 2.6 2.4 2.2 2.0 1.8 1.6 1.4 1.2 1.0 0.8 0.6 0.4 0.2 0.0 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 1.972 1.993 1.975 1.932 1.901 1.938 1.952 1.958 2.001 2.058 2.063 2.009 1.946 1.966 1.973 1.965 1.974 2.014 2.039 1.991 1.93 8 1.942 1.947 1.952 1.967 1.997 2.004 1.932 1.872 1.912 2.222 3.005 4.318 6.145 9.180 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 (Ft.)_Axial Point Unrodded Fxy Core Height (Ft.)Axial Point Unrodded Fxy 14.0 13.8 13.6 13.4 13.2 13.0 12.8 12.6 12.4 12.2 12.0 11.8 11.6 11.4 11.2 11.0 10.8 10,6 10.4 10.2 10.0 9.8 9.6 9.4 9.2 9.0 8.8 8.6 8.4 8.2 8.0 7.8 7.6 7.4 7.2 7.0 1 3 4 5 6 7 8 9 t0 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 6.495 5.203 3.911 2.790 2.552 2.305 2.160 2.087 2.027 2.021*2.030 2.056 2.082 2.074 2,044 2.009 2.038 2.046 2.053 2.082 2.127 2.145 2.113 2.088 2.102 2.110 2,114 2.124 2.161 2.194 2.157 2.124 2.126 2.137 2.148 2.160 6.8, 6.6 6.4 6.2 6.0 5.8 5.6 5.4 5.2 5.0 4.8 4.6 4.4 4.2 4.0 3.8 3.6 3.4 3.2 3.0 2.8 2.6 2.4 2.2 2.0 1.8 1.6 1.4 1.2 1.0 0.8 0.6 0.4 0.2 0.0 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 2.203 2.238 2.204 2.147 2.114 2,106 2.095 2.08 1 2.098 2.131 2.125 2.067 2.02 1 2,016 2.005 1.990 1,992 2.026 2.048 1.990 1,935 1.905 1.875 1.864 1.880 1.926 1.954 1.932 1.945 2.054 2.4 18 3.143 4.250 5.782 8.469 Nuclear Operating Company South Texas Prolect Electric 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 1 RE1 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: (paper copy)Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 1600 East Lamar Boulevard Arlington, TX 76011-4511 Lisa M. Regner Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North (O8H04)11555 Rockville Pike Rockville, MD 20852 NRC Resident Inspector U. S. Nuclear Regulatory Commission P. O. Box 289, Mail Code: MNl16 Wadsworth, TX 77483 (electronic copy)Morgqan. Lewis & Bockius LLP Steve Frantz, Esquire U.S. Nuclear Regulatory Commission Lisa M. Regner NRG South Texas LP John Ragan Chris O'Hara Jim von Suskil CPS Enerqy Kevin Polio Cris Eugster L. D. Blaylock 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~ 1 Cycle 20 Nuclear operating Company Core Operating Limits Report Rev. 1,.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 yle Op 20in Limilts 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 K 1 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'_ 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, =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, ¶-7 = 1 0 sec 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 K 6 Overpower AT reactor trip setpoint Tavg heatup coefficient Ks = 0.002/°F for T>T".,andKQ O forT_< T" T" Indicated full power T"_< 592.0 0 F f,_(AI) =0 for all (Al) lT M Unit 1 Cycle 20 Nucl Operating Company 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.
.'omp..y CUnitOperating Limits Report RV Opert n Cm-"y Core5 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 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 F 0 (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 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 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= 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~ 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 3 A.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 (%)
Unit 1 Cycle 20 Lmt eotRv Nucler Opratig CopanyCore Operating Lmt eotRv m r Page 9 of 16 Figure 2 Required Shutdown Margin for Modes 3 & 4 7.0 6.0 5.0..~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 Unt Cycle 20 Nuclea Oprtn Copn Core Operating Limits Report Rev. I Pagl 10 of 16 Figure 3 Required Shutdown Margin for Mode 5 7.0 6.0 5.0= 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. 1r Page t11 of" 6 Figure 4 MTC versus Power Level 7.0 6.0 5.0 p"4.0 I,d S2.0~0.0 Unacceptable Acceptable1
-1.0-2.0-3.0 0 10 20 30 40 50 60 70 80 90 100 Rated Thermal.Powser
(%)
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 (i2"3,-2-59):-
122-Step ...Overlap H(21,254):
117 Step Overlap 1]
I-4 _(79,259);-A( 77,254 122 Step Overlap 117 Step Overlap I 1 I I I I .A 'l I , .., ; t I I I I I I 1J I I ,-240 220 200 180 160140 S120 S100 80 60 40 20 0 I (,,65 Coto~akAisarad ihdantoulOu o itin 0 10 20 30 40 50 60 Rated Thermal Power (%)70 80 90 100 Nuclear Operating Company UniteIOCyclei20 Limits Report Rex'. 1 Coe peatngPage 1 3of1 Figure 6 AFD Limits versus Power Level 120 110 100 90 80 I-70 60 50 40 30 20 10 0 , , , : I 1_ .2K ---9oI... .. , I _ _ I,9o I.7- ....(-31 ,5 ) 31 ,50)-50-40 20 -10 0 10 20 30 40 50 Axial Flux Difference
(% Delta-I)
SlmT ~iUnit I Cycle 20 Nuclear Operating Company Core Operating Limits Report Rev. 1 Page 14 of 16 1.2!.i 1.0 0.9 0.8 0.0.6~0.4 0.3 0.2 0.1 0.0 Figure 7 K(Z) -Normalized FQ(Z) versus Core Height I .. I'5 .2 i............
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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 (.Ft.)Axial Point Unrodded Fxy Core Height (Ft.)Axial Point Unrodded Fxy 14.0 13.8 13.6 13.4 13.2.13.0 12.8 12.6 12.4 12.2 12.0 11.8 11.6 11.4 11.2 11.0 10.8 10.6 10.4 10.2 10.0 9.8 9.6 9.4 9.2 9.0 8.8 8.6 8.4 8.2 8.0 7.8 7.6 7.4 7.2 7.0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28.29 30 31 32 33 34 35 36 7.3 17 5.739 4.158 2.789 2.495 2.226 2.138 2.118 2.068 2.022 2.0 02 2.014 2.037 2.014 1.965 1.933 1.925 1.920 1.918 1.938 1.973 1.981 i1.943 1.908 1.904 1.895 1.896 1.9 16 1.983 2.036 1.977 1.925 1.929 1.945 1.947 1.939 6.8 6.6 6.4 6.2 6.0 5.8 5.6 5.4 5.2 5.0 4.8 4.6 4.4 4.2 4.0 3.8 3.6 3.4 3.2 3.0 2.8 2.6 2.4 2.2 2.0 1.8 1.6 1.4 1.2 1.0 0.8 0.6 0.4 0.2 0.0 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 1.972 1.993 1.975 1.932 1.901 1.938 1.952 1.958 2.001 2.058 2.063 2.009 1.946 1.966 1.973 1.965 1.974 2.014 2.039 1.991 1.93 8 1.942 1.947 1.952 1.967 1.997 2.004 1.932 1.872 1.912 2.222 3.005 4.318 6.145 9.180 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 (Ft.)_Axial Point Unrodded Fxy Core Height (Ft.)Axial Point Unrodded Fxy 14.0 13.8 13.6 13.4 13.2 13.0 12.8 12.6 12.4 12.2 12.0 11.8 11.6 11.4 11.2 11.0 10.8 10,6 10.4 10.2 10.0 9.8 9.6 9.4 9.2 9.0 8.8 8.6 8.4 8.2 8.0 7.8 7.6 7.4 7.2 7.0 1 3 4 5 6 7 8 9 t0 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 6.495 5.203 3.911 2.790 2.552 2.305 2.160 2.087 2.027 2.021*2.030 2.056 2.082 2.074 2,044 2.009 2.038 2.046 2.053 2.082 2.127 2.145 2.113 2.088 2.102 2.110 2,114 2.124 2.161 2.194 2.157 2.124 2.126 2.137 2.148 2.160 6.8, 6.6 6.4 6.2 6.0 5.8 5.6 5.4 5.2 5.0 4.8 4.6 4.4 4.2 4.0 3.8 3.6 3.4 3.2 3.0 2.8 2.6 2.4 2.2 2.0 1.8 1.6 1.4 1.2 1.0 0.8 0.6 0.4 0.2 0.0 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 2.203 2.238 2.204 2.147 2.114 2,106 2.095 2.08 1 2.098 2.131 2.125 2.067 2.02 1 2,016 2.005 1.990 1,992 2.026 2.048 1.990 1,935 1.905 1.875 1.864 1.880 1.926 1.954 1.932 1.945 2.054 2.4 18 3.143 4.250 5.782 8.469