NOC-AE-07002179, Cycle 13 Core Operating Limits Report

Cycle 13 Core Operating Limits Report

ML071770128
Person / Time
Site:	South Texas
Issue date:	06/20/2007
From:	Leazar D South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-07002179, STI: 32174481
Download: ML071770128 (18)

Text

4t Nuclear Operating Company South Temas Proed Electric GeneratingStation PO. Box289 Wadsworth, Tixas 77483 -

June 20, 2007 NOC-AE-07002179 10CFR50.36 STI: 32174481 U. S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738 South Texas Project Unit 2 Docket No. STN 50-499 Unit 2 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 2 Cycle 13. This report reflects core design changes made during the 2RE12 refueling outage.

There are no commitments in this letter.

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 Manager, Fuels & Analysis mk

Attachment:

Unit 2 Cycle 13 Core Operating Limits Report, Rev. 0.

NOC-AE-07002179 Page 2 of 2 cc:

(paper copy) (electronic copy)

Regional Administrator, Region IV A. H. Gutterman, Esquire U. S. Nuclear Regulatory Commission Morgan, Lewis & Bockius LLP 611 Ryan Plaza Drive, Suite 400 Arlington, Texas 76011-8064 Mohan C. Thadani U. S. Nuclear Regulatory Commission Mohan C. Thadani Thad Hill Project Manager Eddy Daniels U.S. Nuclear Regulatory Commission Marty Ryan One White Flint North (MS 7 D1) Harry Holloway 11555 Rockville Pike Steve Winn Rockville, MD 20852 NRG South Texas LP Senior Resident Inspector Ed Alarcon U. S. Nuclear Regulatory Commission J. J. Nesrsta P. O. Box 289, Mail Code: MN116 R. K. Temple Wadsworth, TX 77483 Kevin Polio City Public Service C. M. Canady Jon C. Wood City of Austin Cox Smith Matthews Electric Utility Department 721 Barton Springs Road C. Kirksey Austin, TX 78704 City of Austin Richard A. Ratliff Bureau of Radiation Control Texas Department of State Health Services 1100 West 49th Street Austin, TX 78756-3189

LM=AjrMM Nuclear Operating Company SOUTH TEXAS PROJECT Unit 2 Cycle 13 CORE OPERATING LIMITS REPORT Revision 0 Core Operating Limits Report Page I of 16

Unit 2 Cycle 13 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 2 of 16 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report for STPEGS Unit 2 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 (Tavg) 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.

M53 Nuclear Operating Company Unit 2 Cycle 13 Core Operating Limits Report Rev. 0 Page 3 of 16 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, T" = 8 sec T2 measured reactor vessel AT lead/lag time constant, T2 = 3 sec T3 measured reactor vessel AT lag time constant, T3 = 2 sec T4 measured reactor vessel average temperature lead/lag time constant, ' 4 = 28 sec T5 measured reactor vessel average temperature lead/lag time constant, Tr5 = 4 sec T6 measured reactor vessel average temperature lag time constant, T6 = 2 sec KI Overtemperature AT reactor trip setpoint, KI = 1.14 K2 Overtemperature AT reactor trip setpoint Tavg coefficient, K2 = 0.028/°F K3 Overtemperature AT reactor trip setpoint pressure coefficient, K 3 = 0.00143/psig T' Nominal full power Tavg, T'< 592.0 OF 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 q, - qb between -70% and +8%, fl(AI) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and q, + qb is total THERMAL POWER in percent of RATED THERMAL POWER; (2) For each percent that the magnitude of q,- 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 q, - 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 T* measured reactor vessel AT lead/lag time constant, "c 1 = 8 sec T2 measured reactor vessel AT lead/lag time constant, T 2 = 3 sec T3 measured reactor vessel AT lag time constant, "c 3 = 2 sec T6 measured reactor vessel average temperature lag time constant, 'r6 = 2 sec T7 Time constant utilized in the rate-lag compensator for Tavg, Tr7 = 10 sec K4 Overpower AT reactor trip setpoint, K4 = 1.08 K5 Overpower AT reactor trip setpoint Tavg rate/lag coefficient, K5 = 0.02/°F for increasing average temperature, and K 5 = 0 for decreasing average temperature K6 Overpower AT reactor trip setpoint Tavg heatup coefficient K6 = 0.002/'F for T> T",andK 6 = 0forT<__ T" T" Indicated full power Tavg, T":5 592.0 OF f2(AI) = 0 for all (AI)

Unit 2 Cycle 13 N A ngACmy Core Operating Limits Report Rev. 0 I EPage 4 of 16 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 pcm/°F.

2.4.3 The 300 ppm, ARO, HFP, MTC shall be less negative than -53.6 pcm/nF (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 pcm/°F 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 either 257 steps withdrawn 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 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).

Unit 2 Cycle 13 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 5 of 16 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 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 FQTP = 2.55.

2.7.2 K(Z) is provided in Figure 7.

2.7.3 The Fxy limits for RATED THERMAL POWER (FX!") within specific core planes shall be:

2.7.3.1 Less than or equal to 2.102 for all cycle bumups 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.

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

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

UQ = Uncertainty for power peaking factor as defined in Equation 5-19 of Reference 3.6 UE = Engineering uncertainty factor of 1.03.

This uncertainty is calculated and applied automatically by the BEACON computer code.

I Unit 2 Cycle 13 Nuclear Operating Company Core Operating Limits Report Rev. 0 A AV-Page 6 of 16 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*UE Where:

UQU = Base FQ 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' = 1.62' 2.8.2 PFA = 0.3 2.8.3 Core Power Distribution Measurement Uncertainty for the Enthalpy Rise Hot Channel Factor 2.8.3.1 If the Power Distribution Monitoring System (PDMS) is operable, as defined in the Technical Requirements Manual, the core power distribution measurement uncertainty (UFn) to be applied to the RH using the PDMS shall be calculated by:

UFAH = 1.0 + (UA/100)

Where:

Um = Uncertainty for power peaking factor as defined in Equation 5-19 of Reference 3.6 This uncertainty is calculated and applied automatically by the BEACON computer code.

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

UFAH= 1.04 Applies to all fuel in the Unit 2 Cycle 13 Core.

• [pAI*" Unit 2 Cycle 13 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 7 of 16

.2.9 DNB PARAMETERS (Specification 3.2.5):

2.9.1 The following DNB-related parameters shall be maintained within the following limits:1 2.9.1.1 Reactor Coolant System Tavg :s 595 'F 2 2.9.1.2 Pressurizer Pressure > 2200 psig 3, 2.9.1.3 Minimum Measured Reactor Coolant System Flow > 403,000 gpm 4 .

3.0 REFERENCES

3.1 Letter from J. D. O'Hare (Westinghouse) to D. F. Hoppes (STPNOC), "South Texas Project Nuclear Operating Company South Texas Project Electric Generating Station Unit 2 Cycle 13 Reload Evaluation (RE)," ST-UB-NOC-07002736, February 1, 2007.

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. 4, "Loop Uncertainty Calculation for Narrow Range Pressurizer Pressure Monitoring Instrumentation," Section 2.3, Page 9.

3.5 Condition Report Engineering Evaluation 05-15241-9, Revision 0, "Unit 2 Cycle 13 Reload Safety Evaluation and Core Operating Limits Report Modes 1, 2, 3, 4, and 5."

3.6 WCAP-12472-P-A, BEACON Core Monitoring and Operations Support System, August 1994.

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

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

3 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.

4 Includes a 2.8% flow measurement uncertainty.

dF*l AMl Unit 2 Cycle 13 Nuclear Operating Company Core Operating Limits Report Rev. 0

ý .AW M-" Page 8 of 16 Figure 1 Reactor Core Safety Limits - Four Loops in Operation 680 660 640 620 600 580 560 540 0 20 40 60 80 100 120 140 Rated Thermal Power (%)

Unit 2 Cycle 13 Nuclear Operating Company Core Operating Limits Report Rev. 0 i-EA A-- Page 9 of 16 Figure 2 Required Shutdown Margin for Modes 3 & 4 7.0 6.0 5.0 4.0 I-3.0 2.0 1.0 0.0 0 400 800 1200 1600 2000 2400 RCS Critical Boron Concentration (ppm)

(for ARI minus most reactive stuck rod)

Unit 2 Cycle 13 Nuclear Operatin'g Company Core Operating Limits Report Rev. 0

-E NA-- Page 10 of 16 Figure 3 Required Shutdown Margin for Mode 5 7.0 6.0 5.0 4.0 S3.0

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

(for ARI minus most reactive stuck rod)

Nuclear Operating oman Unit 2 Cycle 13 triimAFA-

-EMo-mp Core Operating Limits Report Rev. 0 Page 11 of 16 Figure 4 MTC versus Power Level 7.0 6.0 5.0 4.0 0

3.0 2

2.0 0

1.0 CU 0.0

-1.0

-2.0

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

Unit 2 Cycle 13

-ENB Nuclear Operating Company

-- Core Operating Limits Report Rev. 0 Page 12 of 16 Figure 5 Control Rod Insertion Limits* versus Power Level 260 (23,259): 122 Step Overlap - - 79,259): 122Step Overlap

....- (22,257): 120StepOverlapý 'OF 78,257): 120StepOverlap 240 Ba' 220 200 F(0, 202 ) -0 000,74 -

180 Bank C _ oe S160 -----

P on Cid

10. F w a l t d w s o c b r h i e 5 5 rA

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.withdrawn J 10 2304 0

=120 RtdTemlPwr(

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010 20 30 40 50 60 70 80 90 100 Rated Thermal Power M%

Unit 2 Cycle 13 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 13 of 16 Figure 6 AFD Limits versus Power Level 120 110 100 90 80

-- Unacceptable - I nacceptable Operation Acpal.-Operation 70 I-0 Cu 60 LLLL

= I_=LýI 50 Cu (-31,50) --- F5,0 40 30 20 10 0

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

i rU~

Unit 2 Cycle 13 Nula eaigCmpany Core Operating Limits Report Rev. 0 Page 14 of 16 Figure 7 K(Z) - Nornalized FQ(Z) versus Core Height 1.2 L

1.1 1.0 0.9 H.1. 111.

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0 0.4

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1 . I. I I . 1 1 4--

0.0 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 Core Height (ft)

Unit 2 Cycle 13 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 15 of 16 Table 1 (Part 1 of 2)

Unrodded Fxy for Each Core Height for Cycle Burnups Less Than 10000 MWD/MTU Core Height Axial Unrodded Core Height Axial Unrodded (Ft.) Point Fxy (Ft.) Point Fxy 14.00 1 5.922 6.80 37 1.918 13.80 2 4.917 6.60 38 1.911 13.60 3 3.913 6.40 39 1.904 13.40 4 2.909 6.20 40 1.898 13.20 5 2.450 6.00 41 1.896 13.00 6 2.234 5.80 42 1.897 12.80 7 2.191 5.60 43 1.900 12.60 8 2.172 5.40 44 1.904 12.40 9 2.148 5.20 45 1.909 12.20 10 2.106 5.00 46 1.917 12.00 11 2.066 4.80 47 1.926 11.80 12 2.039 4.60 48 1.937 11.60 13 2.032 4.40 49 1.944 11.40 14 2.029 4.20 50 1.948 11.20 15 2.022 4.00 51 1.950 11.00 16 2.012 3.80 52 1.947 10.80 17 2.002 3.60 53 1.944 10.60 18 1.995 3.40 54 1.942 10.40 19 1.989 3.20 55 1.941 10.20 20 1.987 3.00 56 1.941 10.00 21 1.986 2.80 57 1.937 9.80 22 1.986 2.60 58 1.914 9.60 23 1.988 2.40 59 1.882 9.40 24 1.990 2.20 60 1.840 9.20 25 1.995 2.00 61 1.792 9.00 26 2.003 1.80 62 1.744 8.80 27 2.013 1.60 63 1.698 8.60 28 2.024 1.40 64 1.676 8.40 29 2.034 1.20 65 1.668 8.20 30 2.042 1.00 66 1.678 8.00 31 2.044 0.80 67 1.745 7.80 32 2.042 0.60 68 1.936 7.60 33 2.024 0.40 69 2.187 7.40 34 1.996 0.20 70 2.439 7.20 35 1.996 0.00 71 2.690 7.00 36 1.936

AP Unit 2 Cycle 13 Nuclear Operating Company Core Operating Limits Report Rev. 0 WEEM- Page 16 of 16 Table 1 (Part 2 of 2)

Unrodded Fxy for Each Core Height for Cycle Burnups Greater Than or Equal to 10000 MWD/MTU Core Height Axial Unrodded Core Height Axial Unrodded (Ft.) Point Fxy (Ft.) Point Fxy 14.00 1 5.665 6.80 37 2.155 13.80 2 4.742 6.60 38 2.150 13.60 3 3.819 6.40 39 2.138 13.40 4 2.896 6.20 40 2.124 13.20 5 2.455 6.00 41 2.111 13.00 6 2.229 5.80 42 2.099 12.80 7 2.153 5.60 43 2.088 12.60 8 2.105 5.40 44 2.076 12.40 9 2.069 5.20 45 2.065 12.20 10 2.034 5.00 46 2.054 12.00 11 2.006 4.80 47 2.044 11.80 12 2.009 4.60 48 2.034 11.60 13 2.012 4.40 49 2.023 11.40 14 2.018 4.20 50 2.011 11.20 15 2.026 4.00 51 1.999 11.00 16 2.033 3.80 52 1.987 10.80 17 2.036 3.60 53 1.976 10.60 18 2.040 3.40 54 1.965 10.40 19 2.043 3.20 55 1.952 10.20 20 2.047 3.00 56 1.938 10.00 21 2.053 2.80 57 1.920 9.80 22 2.059 2.60 58 1.889 9.60 23 2.065 2.40 59 1.854 9.40 24 2.072 2.20 60 1.839 9.20 25 2.078 2.00 61 1.833 9.00 26 2.083 1.80 62 1.829 8.80 27 2.088 1.60 63 1.829 8.60 28 2.093 1.40 64 1.833 8.40 29 2.098 1.20 65 1.872 8.20 30 2.105 1.00 66 1.960 8.00 31 2.113 0.80 67 2.201 7.80 32 2.122 0.60 68 2.616 7.60 33 2.131 0.40 69 3.119 7.40 34 2.140 0.20 70 3.621.

7.20 35 2.149 0.00 71 4.123 7.00 36 2.154