NOC-AE-05001873, Cycle 13 Core Operating Limits Report

Cycle 13 Core Operating Limits Report

ML051160267
Person / Time
Site:	South Texas
Issue date:	04/20/2005
From:	Leazar D South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-05001873, STI:31875234
Download: ML051160267 (17)

Text

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&W289 Hfdsmwi Toxas 77483 AA April 20, 2005 NOC-AE-05001873 STI: 31875234 10CFR50.36 U. S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852 South Texas Project Unit 1 Docket Nos. STN 50-498 Unit 1 Cycle 13 Core Operating Limits Report In accordance with Technical Specification 6.9.1.6.d, the attached Core Operating Limits Report is submitted for Unit 1 Cycle 13. This report reflects core design changes made during the"1 REX 2 refueling outage.

If there are any questions concerning this report, please contact Scott Head at (361) 972-7136 or me at (361) 972-7795.

David A. Leazar Director, Nuclear Fuel & Analysis jal

Attachment:

1. Unit 1 Cycle 13 Core Operating Limits Report, Rev. 0.

NOC-AE-05001873 Page 2 of 2 cc:

(paper copy) (electronic copy)

Bruce S. Mallett A. H. Gutterman, Esquire Regional Administrator, Region IV Morgan, Lewis & Bockius LLP U. S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 David H. Jaffe Arlington, Texas 76011-8064 U. S. Nuclear Regulatory Commission U. S. Nuclear Regulatory Commission Jack A. Fusco Attention: Document Control Desk Michael A. Reed One White Flint North Texas Genco, LP 11555 Rockville Pike Rockville, MD 20852 C. A. Johnson AEP Texas Central Company Richard A. Ratliff C. Kirksey Bureau of Radiation Control City of Austin Texas Department of State Health Services 1100 West 49th Street Jon C. Wood Austin, TX 78756-3189 Cox Smith Matthews Jeffrey Cruz J. J. Nesrsta U. S. Nuclear Regulatory Commission R. K. Temple P. 0. Box 289, Mail Code: MN116 E. Alarcon Wadsworth, TX 77483 City Public Service C. M. Canady City of Austin Electric Utility Department 721 Barton Springs Road Austin, TX 78704

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Nuclear Operating Company SOUTH TEXAS PROJECT Unit 1 Cycle 13 CORE OPERATING LIMITS REPORT Revision 0 Core Operating Limits Report Page I of 15

'- rA Unit 1Cycle 13 Nuclear Operating Company Core Operating Limits Report Rev. 0 UN AV AVPage 2 of 15

• '.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report for STPEGS Unit 1 Cycle 13 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.

The Technical Specifications affected by this report are:

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) 3/4.1.3.5 SHUTDOWN ROD INSERTION LIMITS

6) 3/4.1.3.6 CONTROL ROD INSERTION LIMITS

7) 3/4.2.1 AFD 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 presented 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 (Tad shall not exceed the limits shown in Figure 1.

2.2 LIMITING SAFETY SYSTEM SETTINGS (Specification 2.2):

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

N ErOrA Unit 1Cycle 13 NuiclearOperating Company Core Operating Limits Report Re 3 Page 3 of 15 2.2.2 The Over-temperature AT and Over-power AT setpoint parameter values are listed below:

Over-temperature AT Setpoint Parameter Values T. measured reactor vessel AT lead/lag time constant, t1 = 8 sec t2 measured reactor vessel AT lead/lag time constant, r2 = 3 sec

¶3 measured reactor vessel AT lag time constant, r3 = 2 sec

¶4 measured reactor vessel average temperature lead/lag time constant, T4 = 28 sec T5 measured reactor vessel average temperature lead/lag time constant, T5 = 4 sec T6 measured reactor vessel average temperature lag time constant, T6 = 2 sec K, Overtemperature AT reactor trip setpoint, K, = 1.14 K2 Overtemperature AT reactor trip setpoint Ta, coefficient, K2 = 0.028/1F K3 Overtemperature AT reactor trip setpoint pressure coefficient, K3 = 0.00 I43/psig T' Nominal full power Tavg, T'c 592.0 IF P' Nominal RCS pressure, P' = 2235 psig f1(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 qt - qb between -70% and +8%, fl(AT) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + qb is total THERMAL POWER in percent of RATED THERMAL POWER; (2) For each percent that the magnitude of qt - qb 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.

Over-power AT Setpoint Parameter Values Tl measured reactor vessel AT lead/lag time constant, Tr = 8 sec

-2 measured reactor vessel AT lead/lag time constant, T2 = 3 sec T3 measured reactor vessel AT lag time constant, T3 = 2 sec T6 measured reactor vessel average temperature lag time constant, ¶5 = 2 sec T, Time constant utilized in the rate-lag compensator for Tavg, ¶7 = 10 sec K4 Overpower AT reactor trip setpoint, K4 = 1.08 K5 Overpower AT reactor trip setpoint Tavg rate/lag coefficient, Ks = 0.02/°F for increasing average temperature, and K5 = 0 for decreasing average temperature K6 Overpower AT reactor trip setpoint Tavg heatup coefficient K6 = 0.002/°F for T>T",and K6 = 0forT< Tr T" Indicated full power Tavg, r<5 592.0 °F f 2(AI) = 0 for all (AI)

- T M Unit 1Cycle 13 Nuclear Operating Company Core Operating Limits Report Rev. 0 4W N AFPage 4 of 15 2.3 SHUTDOWN MARGIN (Specification 3.1.1.1):

The SHUTDOWN MARGIN 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 pcmI0 F.

2.4.3 The 300 ppm, ARO, HFP, MTC shall be less negative than -53.6 pcm/IF (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, HFP stands for Hot Full Power (100% RATED THERMAL POWER),

HFP vessel average temperature is 592 'F.

2.4.4 The Revised Predicted near-EOL 300 ppm MTC shall be calculated using the algorithm from T.S. 6.9.1.6.b.10:

Revised Predicted MTC = Predicted MTC + AFD Correction - 3 pcni/0F If the Revised Predicted MTC is less negative than the S.R. 4.1.1.3b limit and all of the benchmark 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 at least 250 steps withdrawn but not exceeding 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 withdrawn 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).

• GUnit 1 Cycle 13 Nuclear Operating Company Core Operating Limits Report Rev. 0 Uff W AVPage 5 of 15 2.6 AXIAL FLUX DIFFERENCE (Specification 3.2.1):

2.6.1 AFD limits as required by Technical Specification 3.2.1 are determined by 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 FR'rP = 2.55.

2.7.2 K(Z) is provided in Figure 7.

2.7.3 The F.Y limits for RATED THERMAL POWER (FR;) within specific core planes shall be:

2.7.3.1 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 FXY limits were used to confirm that the heat flux hot channel factor FQ(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-8385. Therefore, these FXY limits provide assurance that the initial conditions assumed in the LOCA analysis are met, along with the ECCS acceptance criteria of 10 CFR 50.46.

APWARNUnit 1Cycle 13 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 6 of 15 2.8 ENTHALPY RISE HOT CHANNEL FACTOR (Specification 3.2.3):

2.8.1 FRT' = 1.557' 2.8.2 PFA,1 = 0.3 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 Tavg, < 595 'F ,

2.9.1.2 Pressurizer Pressure, > 2200 psig4 ,

2.9.1.3 Minimum Measured Reactor Coolant System Flow 5 > 403,000 gpm.

3.0 REFERENCES

3.1 Letter from D. E. Robinson (Westinghouse) to D. F. Hoppes (STPNOC), "Unit I Cycle 13 Final Reload Evaluation (RE) Revision 1," ST-UB-NOC-05002532, Rev. 1, March 29 2005.

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

3.3 STPNOC Calculation ZC-7035, Rev. 2, "Loop Uncertainty Calculation for RCS Tavg Instrumentation," Section 10.1, effective July 22, 2003.

3.4 STPNOC Calculation ZC-7032, Rev. 4, "Loop Uncertainty Calculation for Narrow Range Pressurizer Pressure Monitoring Instrumentation," Section 2.3, Page 9, effective July 22, 2003.

3.5 Condition Report Engineering Evaluation 03-6461-9, Revision 0, "Reload Safety Evaluation and Core Operating Limits Report for South Texas Unit I Cycle 13 Modes 1, 2, 3, 4, and 5."

I Applies to all fuel in the Unit I Cycle 13 Core.

2 A discussion of the processes to be used to take these readings is provided in the basis for Technical Specification 3.2.5.

3 Includes a 1.9 'F measurement uncertainty per Reference 3.3.

4 Limit not applicable during either a Thermal Power ramp in excess of 5% of RTP per minute or a Thermal Power step in excess of 10% RTP. Includes a 9.6 PSI measurement uncertainty as read on QDPS display per Reference 3.4.

5 Includes a 2.8% flow measurement uncertainty.

Unit 1 Cycle 13 NuclAr OperatingCompany Core Operating Limits Report Rev. 0 f ND- Page 7 of 15 Figure 1 Reactor Core Safety Limits - Four Loops in Operation 680 660 640 620 0EL U:

600 580 560 540 0 20 40 60 80 100 120 140 Rated Ihermal Power (%)

Unit 1 Cycle 13 Nuclear Operating Company Core Operating Limits Report Rev. 0 Na- Page 8 of 15 Figure 2 Required Shutdovin Margin for Modes 3 & 4 7.0 I 1 I I I I I 6.0 I 1  ! II _ 1 1 1 -I I I_

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(for ARI minus most reactive stuck rod)

Unit 1 Cycle 13

• PWA Nuclear Operating Company Core Operating Limits Report Rev. 0 ONN-Page 9 of 15 Figure 3 Required Shutdoivn Margin for Mode 5

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Unit 1 Cycle 13 Nuclear Operating Company Core Operating Limits Report Rev. 0 D - Page 10 of 15 Figure 4 MTC versus Powver Level I T I I FI iI 7.0 6.0 IlI 1 Unacceptable OperationI -

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Nucea Ot 60N A-Unit 1 Cycle 13 Core Operating Limits Report Rev. 0 Page 11 of 15 Figure 5 Control Rod Insertion Limits* versus Power Level (23 259): 122 Step Overlap ( 79, 259 ): 122 Step Overlap (23 258): 121 Step Overlap ( 79, 258 ): 121 Step Overlapi 260 I 22 256): 119 Step Overlap . . I . . . . . M . I . . .I I 0 78, 256 ): 119 Step Overlap!

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MM~w Unit 1 Cycle 13 Core Operating Limits Report Rev. 0 EfNula Apratn -opn Page 12 of 15 Figure 6 AFD Limits versus Power Level 120 110 I 1:1I T1z11 il ILI T XX t I I1l f 1 T lr1 I rII II I Il 100 I I l I _ l I II T V_ i I II _I lI II I \I I i I I I I I _

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d l M Unit 1Cycle 13 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 13 of 15 Figure 7 K(Z) - Normalized FQ(Z) versus Core Height 1.2 1.1

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Unit 1 Cycle 13 Nuclear Operating Company Core Operating Limits Report Rev. 0 4W AV K -Page 14 of 15 Table 1 (Part 1 of 2)

Unrodded F1 Yfor Each Core Height for Cycle Burnups Less Than 10500 MLWD/MTU Core Height Axial Unrodded Core Height Axial Unrodded (Ft.) Point Fxy 4 (Ft.)

Point Fxy 14.00 1 4.633 6.80 37 1.872 13.80 2 3.954 6.60 38 1.861 13.60 3 3.275 6.40 39 1.853 13.40 4 2.596 6.20 40 1.845 13.20 5 2.276 6.00 41 1.841 13.00 6 2.068 5.80 42 1.840 12.80 7 2.070 5.60 43 1.839 12.60 8 2.046 5.40 44 1.840 12.40 9 2.023 5.20 45 1.844 12.20 10 1.995 5.00 46 1.850 12.00 11 1.979 4.80 47 1.860 11.80 12 1.971 4.60 48 1.870 11.60 13 1.971 4.40 49 1.878 11.40 14 1.975 4.20 50 1.885 11.20 15 1.979 4.00 51 1.890 11.00 16 1.984 3.80 52 1.893 10.80 17 1.984 3.60 53 1.893 10.60 18 1.984 3.40 54 1.895 10.40 19 1.982 3.20 55 1.899 10.20 20 1.985 3.00 56 1.904 10.00 21 1.987 2.80 57 1.903 9.80 22 1.989 2.60 58 1.911 9.60 23 1.997 2.40 59 1.921 9.40 24 2.003 2.20 60 1.932 9.20 25 2.011 2.00 61 1.929 9.00 26 2.017 1.80 62 1.920 8.80 27 2.023 1.60 63 1.905 8.60 28 2.027 1.40 64 1.907 8.40 29 2.032 1.20 65 1.909 8.20 30 2.037 1.00 66 1.920 8.00 31 2.033 0.80 67 1.994 7.80 32 2.016 0.60 68 2.124 7.60 33 1.983 0.40 69 2.281 7.40 34 1.945 0.20 70 2.439 7.20 35 1.911 0.00 71 2.596 7.00 36 1.889

Unit 1 Cycle 13 Nuclear Operating Company Core Operating Limits Report Rev. 0 or i-Page 15 of 15l Table 1 (Part 2 of 2)

Unrodded Fxy for Each Core Height for Cycle Burnups Greater Than or Equal to 10500 MWD/IMTU Core Height Axial Unrodded Core Height Axial Unrodded (Ft.) Point Fxy (Ft.) Point Fxy 14.00 1 4.778 6.80 37 2.144 13.80 2 4.129 6.60 38 2.139 13.60 3 3.480 6.40 39 2.127 13.40 4 2.831 6.20 40 2.115 13.20 5 2.450 6.00 41 2.101 13.00 6 2.153 5.80 42 2.089 12.80 7 2.149 5.60 43 2.077 12.60 8 2.121 5.40 44 2.067 12.40 9 2.096 5.20 45 2.057 12.20 10 2.055 5.00 46 2.047 12.00 11 2.032 4.80 47 2.038 11.80 12 2.026 4.60 48 2.028 11.60 13 2.019 4.40 49 2.017 11.40 14 2.022 4.20 50 2.006 11.20 15 2.027 4.00 51 1.994 11.00 16 2.031 3.80 52 1.982 10.80 17 2.033 3.60 53 1.969.

10.60 18 2.035 3.40 54 1.957 10.40 19 2.036 3.20 55 1.945 10.20 20 2.041 3.00 56 1.931 10.00 21 2.048 2.80 57 1.916 9.80 22 2.056 2.60 58 1.891 9.60 23 2.065 2.40 59 1.858 9.40 24 2.072 2.20 60 1.841 9.20 25 2.078 2.00 61 1.839 9.00 26 2.084 1.80 62 1.835 8.80 27 2.089 1.60 63 1.836 8.60 28 2.095 1.40 64 1.853 8.40 29 2.100 1.20 65 1.853 8.20 30 2.106 1.00 66 1.877 8.00 31 2.112 0.80 67 2.127 7.80 32 2.119 0.60 68 2.605 7.60 33 2.127 0.40 69 3.195 7.40 34 2.135 0.20 70 3.786 7.20 35 2.142 0.00 71 4.377 7.00 36 2.145