ML081340348

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Submittal of Cycle 15 Core Operating Limits Report
ML081340348
Person / Time
Site: South Texas STP Nuclear Operating Company icon.png
Issue date: 05/07/2008
From: Leazar D
South Texas
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-8002302, STI: 32308506
Download: ML081340348 (18)
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Nuclear Operating Company South Texas ProjectElectric GeneratingStation PO.A*x 289 Wadsworth, Txas 77483 A A-May 7, 2008 NOC-AE-8002302 10CFR50.36 STI: 32308506 U. S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738 South Texas Project Unit 1 Docket No., STN 50-498 Unit 1 Cycle 15 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 15. This report\reflects core design changes made during the 1 RE14 refueling outage.

There are no commitments in this letter.

If there are any questions concerning this report, please contact Marilyn Kistler at (361),

972-8385 or me at (361) 972-7795.

David A. Leaza .

Manager, Nuclear Fuels & Analysis mk

Attachment:

Unit 1 Cycle 15 Core Operating Limits Report, Rev. 0

NOC-AE-08002302 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 Senior Project Manager Eddy Daniels U.S. Nuclear Regulatory Commission Catherine Callaway One White Flint North (MS 7 D1) Brad Porlier 11555 Rockville Pike Steve Winn Rockville, MD 20852 Staney Rostad NRG South Texas LP Senior Resident Inspector Ed Alarcon U. S. Nuclear Regulatory Commission J. J. Nesrsta P. 0. Box 289, Mail Code: MN1 16 R. K. Temple Wadsworth, TX 77483 Kevin Polio City Public Service U. S. Nuclear Regulatory Commission Jon C. Wood Attention: Document Control Desk Cox Smith Matthews One White Flint North 11555 Rockville Pike C. Kirksey Rockville, MD 20852 City of Austin C. M. Canady City of Austin Electric Utility Department 721 Barton Springs Road Austin, TX 78704 Richard A. Ratliff Bureau of Radiation Control Texas Department of State Health Services 1100 West 49th Street Austin, TX 78756-3189

Nuclear Operating Company SOUTH TEXAS PROJECT Unit 1 Cycle 15 CORE OPERATING LIMITS REPORT Revision 0 Core Operating Limits Report Page I of 16

l&IrAM Unit 1Cycle 15 Nuclear Operating Company Core Operating Limits Report Rev. 0 l AF K - Page 2 of 16 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report for STPEGS Unit 1 Cycle 15 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.

  • dlIrf Unit 1Cycle 15 Nuclear Operating Company 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 Tr measured reactor vessel AT lead/lag time constant, "ij= 8 sec T2 measured reactor vessel AT lead/lag time constant, z2 = 3 sec T3 measured reactor vessel AT lag time constant, "3 = 2 sec T4 measured reactor vessel average temperature lead/lag time constant, cr4 = 28 sec T5 measured reactor vessel average temperature lead/lag time constant, t 5 = 4 sec T6 measured reactor vessel average temperature lag time constant, 16 = 2 sec K1 Overtemperature AT reactor trip setpoint, K1 = 1.14 K2 Overtemperature AT reactor trip setpoint Tavg coefficient, K2 = 0.028/°F K3 Overtemperature AT reactor trip setpoint pressure coefficient, K3 = 0.001 43/psig T' Nominal full power Tavg, T'< 592.0 'F P' Nominal RCS pressure, P' = 2235 psig fl(Al) 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%, f1(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 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 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 T1 measured reactor vessel AT lead/lag time constant, Ti = 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 T6 measured reactor vessel average temperature lag time constant, T6 = 2 sec T7 Time constant utilized in the rate-lag compensator for Tavg, 17 = 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 K5 = 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"< 592.0 'F f 2(AI) = 0 for all (Al)

Unit 1 Cycle 15 Nuclear Operating Company Core Operating Limits Report Rev. 0 AW N rPage-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/PF.

2.4.3 The 300 ppm, ARO, HFP, MTC shall be less negative than -53.6 pcm/°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, 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 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).

  • dli 3 Unit 1Cycle 15 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 FQP = 2.55.

2.7.2 K(Z) is provided in Figure 7.

2.7.3 The Fxy limits for RATED THERMAL POWER (FXyT 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.

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 Section 3.3.3.12, 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.

lllIFAN Unit 1Cycle 15 Nuclear Operating Company Core Operating Limits Report Rev. 0 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):

tzRTP1 2.8.1 FAH = 1.62 2.8.2 PFAH = 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 Section 3.3.3.12, the core power distribution measurement uncertainty (UFAH) to be applied to the FN using the PDMS shall be calculated by:

UFAH = 1.0 + (UA/100)

Where:

UAH = 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 (UFAH) shall be:

UFAH = 1.04 I Applies to all fuel in the Unit 1 Cycle 15 Core.

Unit 1 Cycle 15 Nuclear Operating Company Core Operating Limits Report Rev. 0 SW Repor Page 7of 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 -< 595 -F 2, 2.9.1.2 Pressurizer Pressure > 2200 psig 3, 2.9.1.3 Minimum Measured Reactor Coolant System Flow > 403,000 gpm4 .

3.0 REFERENCES

3.1 Letter from D. V. Lockridge (Westinghouse) to D. F. Hoppes (STPNOC), "South Texas Project Electric Generating Station Unit 1 Cycle 15 Final Reload Evaluation," NF-TG-08-9 Revision 1 (ST-UB-NOC-08002832), February 6, 2008.

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, 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 06-13726-9, Revision 0, "Reload Safety Evaluation and Core Operating Limits Report for South Texas Unit 1 Cycle 15 Modes 1, 2, 3, 4, and 5."

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

I 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 applicableduring 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.

d3I Unit 1Cycle 15 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 8 of 16 Figure 1 Reactor Core Safety Limits - Four Loops in Operation 680 660 640 620 0

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

Unit 1 Cycle 15 Nuclear Operating Company Core Operating Limits Report Rev. 0

ý N-Page r 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 ARI minus most reactive stuck rod)

d&IrAM Nuclear Operating Company Unit 1Cycle 15 Core Operating Limits Report Rev. 0 Page 10 of 16 Figure 3 Required Shutdoiwn Margin for Mode 5 7.0 6.0 5.0 4.0,;

~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)

~dp i Unit 1Cycle 15 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 11 of 16 Figure 4 MTC versus Power Level 7.0 6.0

- - ---- ~Unacceptable Op eration ____

5.0 4.0 4cceptab le Operation~

" 3.0 2.0 1.0 0.0

-2.0

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

~d& Unit 1 Cycle 15 Nuclear Operating Company Core Operating Limits Report Rev. 0 PWae2o--

Page 12 of 16 Figure 5 Control Rod Insertion Limits* versus Power Level FF02 260 f( 23 , 259 ): 122 Step Overlap ( 79 ,259): 122 Step Overlap

-( 22,257): 120 Step Overlap ( 78 ,257): 120 Step Overlap 240 B SCBann already w n F O 220 (Fl w w s 200 . . . .. . . . cOo l a . . f

: :e 180  :  :  :  :

withd0,1wn.:

: -  : :00  :  :

" 160 I j Bank C 00-0

;o

. 140 0

0e, 120  ;;;;  ::aa D]

100 80 0" Cnro ak saled withran oul Ou oiin Ful ihransalb h cniinweesudw n

.. . . ... rl an sar t h poito of eithe 5 r 5 tp 60 withrawn 40 20 0

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

Unit 1 Cycle 15 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 13 of 16 Figure 6 AFD Limits versus Power Level 120 110

.100 F(--1 0(11,90) 90 80 Unacceptable - Unacceptable Oiperation* ¢ Acceptable -Operation#

I-70 Oeatperati 0 I II I E

60 zf 1-4 50

(-31 50) 150) 40 30 20 10 0

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

Unit 1 Cycle 15 Nuclear Operating Company Core Operating Limits Report Rev. 0 Srf Page 14 of 16 Figure 7 K(Z) - Normalized FQ(Z) versus Core Height 1 )

1.1

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0.9 0.8 a 0.7

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'1 Core Elev. (ft) FQ K(Z) 0.0 2.55 1.0 0.5 7.0 2.55 1.0 14.0 2.359 0.925 J f .iL . .I . -Ii I-1 i 0.4 Pi ! ! .. .... . . .. .. . ., .

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--I I I I I i i I I i i i I I I i ii i I I I ! i I iI i i i i I 0.1 0.0 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 Core Height (ft)

TMP Unit 1 Cycle 15 Nuclear Operating Company Core Operating Limits Report Rev. 0 1 AF N-- Page 15 of 16 Table 1 (Part I 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 Fxv (Ft.) Point Fxy 14.00 1 5.615 6.80 37 1.992 13.80 2 4.647 6.60 38 1.987 13.60 3 3.679 6.40 39 1.982 13.40 4 2.710 6.20 40 1.980 13.20 5 2.273 6.00 41 1.977 13.00 6 2.073 5.80 42 1.976 12.80 7 2.040 5.60 43 1.978 12.60 8 2.034 5.40 44 1.981 12.40 9 2.026 5.20 45 1.988 12.20 10 2.006 5.00 46 1.991 12.00 11 1.990 4.80 47 1.993 11.80 12 1.982 4.60 48 1.992 11.60 13 1.982 4.40 49 1.989 11.40 14 1.979 4.20 50 1.987 11.20 15 1.970 4.00 51 1.986 11.00 16 1.954 3.80 52 1.983 10.80 17 1.945 3.60 53 1.978' 10.60 18 1.939 3.40 54 1.968 10.40 19 1.937 3.20 55 1.961 10.20 20 1.942 3.00 56 1.955 10.00 21 1.957 2.80 57 1.950 9.80 22 1.977 2.60 58 1.944 9.60 23 1.994 2.40 59 1.934 9.40 24 2.007 2.20 60 1.918 9.20 25 2.018 2.00 61 1.878 9.00 26 2.028 1.80 62 1.830 8.80 27 2.033 1.60 63 1.777 8.60 28 2.038 1.40 64 1.764 8.40 29 2.054 1.20 65 1.753 8.20 30 2.071 1.00 66 1.760 8.00 31 2.075 0.80 67 1.872 7.80 32 2.064 0.60 68 2.074 7.60 33 2.043 0.40 69 2.321 7.40 34 2.025 0.20 70 2.568 7.20 35 2.008 0.00 71 2.815 7.00 36 1.998

Unit 1 Cycle 15 Nuclear Operating Company Core Operating Limits Report Rev. 0 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 5.780 6.80 37 2.160 13.80 2 4.838 6.60 38 2.155 13.60 3 3.896 6.40 39 2.142 13.40 4 2.954 6.20 40 2.128 13.20 5 2.499 6.00 41 2.114 13.00 6 2.263 5.80 42 2.101 12.80 7 2.182 5.60 43 2.089 12.60 8 2.132 5.40 44 2.077 12.40 9 2.089 5.20 45 2.065 12.20 10 2.054 5.00 46 2.054 12.00 11 2.014 4.80 47 2.044 11.80 12 2.007 4.60 48 2.035 11.60 13 2.009 4.40 49 2.024 11.40 14 2.013 4.20 50 2.013 11.20 15 2.014 4.00 51 2.000 11.00 16 2.015 3.80 52 1.989 10.80 17 2.017 3.60 53 1.979 10.60 18 2.021 3.40 54 1.970 10.40 19 2.028 3.20 55 1.956 10.20 20 2.044 3.00 56 1.935 10.00 21 2.064 2.80 57 1.908 9.80 22 2.086 2.60 58 1.896 9.60 23 2.105 2.40 59 1.870 9.40 24 2.118 2.20 60 1.843 9.20 25 2.126 2.00 61 1.836 9.00 26 2.127 1.80 62 1.833 8.80 27 2.126 1.60 63 1.834 8.60 28 2.125 1.40 64 1.839 8.40 29 2.126 1.20 65 1.876 8.20 30 2.127 1.00 66 1.958 8.00 31 S2.129 0.80 67 2.192 7.80 32 2.134 0.60 68 2.602 7.60 33 2.141 0.40 69 3.099 7.40 34 2.149 0.20 70 3.596 7.20 35 2.156 0.00 71 4.093 7.00 36 2.160

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