NOC-AE-11002673, Cycle 17 Core Operating Limits Report

Cycle 17 Core Operating Limits Report

ML11144A194
Person / Time
Site:	South Texas
Issue date:	05/17/2011
From:	Dunn R South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-11002673
Download: ML11144A194 (19)

Text

Nuclear Operating Company South Texas Pro/ect E1ectric GeneratingStation P.O. Boa 289 Wadsworth. Texas 77483 May 17, 2011 NOC-AE-1 1002673 File No.: G25 10 CFR 50.36 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 17 Core Operatingq Limits Report Pursuant to Technical Specification 6.9.1.6.d, STP Nuclear Operating Company submits the attached Core Operating Limits Report for Unit 1 Cycle 17. The report covers the core design changes made during the 1RE16 refueling outage.

There are no commitments included in this report.

If there are any questions on this report, please contact either Philip Walker at (361) 972-8392 or me at (361) 972-7743.

Roland F. Dunn Manager, Nuclear Fuel & Analysis PLW

Attachment:

Unit 1 Cycle 17 Core Operating Limits AOO/

STI: 32862265

NOC-AE-1 1002673 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 Balwant K. Singal U. S. Nuclear Regulatory Commission Balwant K. Singal John Ragan Senior Project Manager Catherine Callaway U.S. Nuclear Regulatory Commission Jim von Suskil One White Flint North (MS 8B1) NRG South Texas LP 11555 Rockville Pike Rockville, MD 20852 Senior Resident Inspector Ed Alarcon U. S. Nuclear Regulatory Commission Kevin Polio P. 0. Box 289, Mail Code: MN116 Richard Pena Wadsworth, TX 77483 City Public Service C. M. Canady Peter Nemeth City of Austin Crain Caton & James, P.C.

Electric Utility Department 721 Barton Springs Road C. Mele Austin, TX 78704 City of Austin Richard A. Ratliff Texas Department of State Health Services Alice Rogers Texas Department of State Health Services

ATTACHMENT South Texas Project Unit I Cycle 17 Core Operating Limits Report

SLM1V" Nuclear Operating Company SOUTH TEXAS PROJECT Unit 1 Cycle 17 CORE OPERATING LIMITS REPORT Revision 0 Core Operating Limits Report Page I of 19

dL=A lFA Unit 1Cycle 17 Nuclear Operating Company Core Operating Limits Report Rev. 0

ý AW rf__ Page 2 of 16 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report for STPEGS Unit 1 Cycle 17 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.

J& A Unit 1 Cycle 17 Nuclear Operating Company Core Operating Limits Report Rev. 0 A1 ffr Page3 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

-11 measured reactor vessel AT lead/lag time constant, T 1, = 8 sec T2 measured reactor vessel AT lead/lag time constant, -r2 = 3 sec

-r3 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 K1 Overtemperature AT reactor trip setpoint, K, = 1.14 K2 Overtemperature AT reactor trip setpoint Tavg coefficient, K2 = 0.028/°F K3 Overtemperature AT reactor trip setpoint pressure coefficient, K3 = 0.00143/psig T' Nominal full power Tavg, T'< 592.0 'F P' Nominal RCS pressure, P' = 2235 psig fl(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(AI) = 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

'TI measured reactor vessel AT lead/lag time constant, 1, = 8 sec T2 measured reactor vessel AT lead/lag time constant, T2 = 3 sec Tr3 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",andK6 = 0forT< T" T" Indicated full power Tavg, T"< 592.0 'F f2 (AI) = 0 for all (Al)

dp i* "

Nuclear Operating Company Unit I Cycle 17 Core Operating Limits Report Rev. 0 Page 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/°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 the document referenced by Technical Specification 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 COLR Section 2.4.3 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 256 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).

AP A Unit 1 Cycle 17 Nuclear Operating Company Core Ope1Ctil g 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 FP = 2.55.

2.7.2 K(Z) is provided in Figure 7.

2.7.3 The Fxy limits for RATED THERMAL POWER (FRTP) 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 I for all unrodded core planes.

2.7.3.3 PFY = 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 FX,,(Z) using the PDMS shall be calculated by:

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

UQ = Uncertainty for power peaking factor as defined in Equation 5-19 from the document referenced by Technical Specification 6.9.1.6.b. 11.

UE = Engineering uncertainty factor of 1.03.

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

A T M Nuclear Operating Company Unit 1Cycle 17 Core Operating Limits Report Rev. 0 1 Nl--" 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):

ERTP 2.8.1 F = 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 N

FAH using the PDMS shall be the greater of:

UFAH = 1.04 OR UFM = 1.0 + (UJA/IO0)

Where:

UAH = Uncertainty for power peaking factor as defined in Equation 5-19 from the document referenced by Technical Specification 6.9.1.6.b.11.

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 (UFA) shall be:

UFAH = 1.04 I Applies to all fuel in the Unit I Cycle 17 Core.

NP nin* Unit 1 Cycle 17 NulArCompany Operating 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:-

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 J. M. Ralston (Westinghouse) to D. F. Hoppes (STPNOC), "South Texas Project Electric Generating Station Unit 1 Cycle 17 Final Reload Evaluation (RE)" NF-TG-1 1-6 (ST-UB-NOC-1 1003139) dated January 25, 2011.

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 09-16959-9, "Unit 1 Cycle 17 Reload Safety Evaluation and Core Operating Limits Report Modes 1, 2, 3, 4, and 5."

3.6 5Z529ZB01025 Rev. 4, Design Basis Document, Technical Specifications /LCO, Tech Spec Section 3.2.5.c 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 the most limiting flow measurement uncertainty of 2.8% from Reference 3.6.

ApIr"* Unit 1 Cycle 17 Nuclear Operating Company Core Operating Limits Report Rev. 0

- ANf- Page 8of 16 Figue 1 Reactor Core Safety Limits - Four Loops in Operation 680 660 640 620 U,.

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

• LA Unit 1Cycle 17 Nuclear Operating Company Core Operating Limits Report Rev. 0 A* ArPage - 9 of 16 Figure 2 Required Shutdown Margin for Modes 3 & 4 7.0 6.0 5.0 S4.0

! 3.0 2.0 (C

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

(for ARI minus most reactive stuck rod)

• qT M Unit 1Cycle 17 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 10 of 16 Figure 3 Required Shutdown 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)

NLrPMnJ Unit 1 Cycle 17 Nuclear Operating Company Core Operating Limits Report Rev. 0

ý AW r- Page 11 of 16 Figue 4 MTC verus Power Level 7.0 6.0 Unacceptable OperationI 5.0 4.0 Acceptable Operation>

3.0 U

2.0 1.0 S0.0 I _____

-1.0

-2.0

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

WRIMM Unit 1 Cycle 17 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 12 of 16 Figure 5 Control Rod Insertion Limits* versus Power Level 260 O( 23 ,259): 122 Step Overlap (79 ,259): 122 Step Overlap (22,256): 119 Step Overlap I I I I I I I I I I I 1 I (78 ,256): 119 Step Overlap OF I: :I: II 11 1 . I01 240 220 0,! 202, 200

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

ATiM" Unit 1 Cycle 17 Nuclear Operating Company Core Operating Limits Report Rev. 0 W A MPage 13 of 16 AFD Limits versus Power Level 120 110 100

-(-11,90) (11,90)-

90 80 IiI 11 Unacceptable q Unacceptable Operation Acceptable: - --- Operation I 70 I '~~iiiii

-- 7 TTT 11JOperation' PC 60 - -

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50 40

(-31,50 - - -I ,0 - - -(3 30 20 10 0

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

Np irp Unit 1 Cycle 17 Nuclear Operating Company Core Operating Limits Report Rev. 0 1 AW ff - Page 14 of 16 Figure 7 K(Z) - Normalized FQ(Z) versus Core Height 1.2 F-vi i i i i ! i i i i i i ! i ! i i i i i i ! ; i i i i ; i i i i i i i i i i i ! ý i i i i i 1 1 1 . . ý ,

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I II I II I I I I I I I I I II II I I I II I I I I I 0.7 "0.6 Core Elev. (ft) FQ K(Z) 0.0 2.55 1.0 7.0 2.55 1.0 0.5 .0 2.359 0.

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0.4 0.3 I C C I CII I I I CI' I IC I IC I I I: II I: I CC CI CI ii I IC I I I C I il 1 I 0.2 i i iCi iIi i i i i i i I I I ,[

I 1I I I I I I CII I I I I i II I I II I III I I I I I I I II I I 0.1 0.0 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 Core Height (ft)

A 1p/ Ir Unit 1 Cycle 17 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 15 of 16 Table 1 (Part 1 of 2)

Unrodded F. 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 4.867 6.80 37 1.986 13.80 2 4.107 6.60 38 1.985 13.60 3 3.347 6.40 39 1.980 13.40 4 2.587 6.20 40 1.978 13.20 5 2.230 6.00 41 1.978 13.00 6 1.999 5.80 42 1.981 12.80 7 1.999 5.60 43 1.984 12.60 8 1.976 5.40 44 1.987 12.40 9 1.960 5.20 45 1.989 12.20 10 1.946 5.00 46 1.991 12.00 11 1.934 4.80 47 1.991 11.80 12 1.926 4.60 48 1.991 11.60 13 1.922 4.40 49 1.990 11.40 14 1.916 4.20 50 1.982 11.20 15 1.907 4.00 51 1.975 11.00 16 1.894 3.80 52 1.968 10.80 17 1.887 3.60 53 1.962 10.60 18 1.882 3.40 54 1.956 10.40 19 1.879 3.20 55 1.955 10.20 20 1.885 3.00 56 1.947 10.00 21 1.895 2.80 57 1.942 9.80 22 1.903 2.60 58 1.943 9.60 23 1.914 2.40 59 1.949 9.40 24 1.924 2.20 60 1.957 9.20 25 1.938 2.00 61 1.937 9.00 26 1.957 1.80 62 1.899 8.80 27 1.982 1.60 63 1.846 8.60 28 2.009 1.40 64 1.823 8.40 29 2.028 1.20 65 1.829 8.20 30 2.043 1.00 66 1.871 8.00 31 2.052 0.80 67 2.036 7.80 32 2.050 0.60 68 2.468 7.60 33 2.026 0.40 69 3.033 7.40 34 2.005 0.20 70 3.598 7.20 35 1.990 0.00 71 4.163 7.00 36 1.987

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Nuclear Operating Company Unit Core 1Operating Cycle 17 Limits Report Rev. 0

ý AW Ar- 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 4.672 6.80 37 2.167 13.80 2 4.098 6.60 38 2.161 13.60 3 3.525 6.40 39 2.146 13.40 4 2.951 6.20 40 2.131 13.20 5 2.577 6.00 41 2.116 13.00 6 2.265 5.80 42 2.104 12.80 7 2.183 5.60 43 2.092 12.60 8 2.112 5.40 44 2.080 12.40 9 2.056 5.20 45 2.069 12.20 10 2.016 5.00 46 2.058 12.00 11 2.011 4.80 47 2.047 11.80 12 2.006 4.60 48 2.036 11.60 13 2.005 4.40 49 2.024 11.40 14 2.009 4.20 50 2.011 11.20 15 2.012 4.00 51 1.998 11.00 16 2.018 3.80 52 1.985 10.80 17 2.022 3.60 53 1.974 10.60 18 2.027 3.40 54 1.964 10.40 19 2.031 3.20 55 1.953 10.20 20 2.046 3.00 56 1.940 10.00 21 2.065 2.80 57 1.926 9.80 22 2.087 2.60 58 1.903 9.60 23 2.106 2.40 59 1.875 9.40 24 2.120 2.20 60 1.852 9.20 25 2.129 2.00 61 1.846 9.00 26 2.130 1.80 62 1.844 8.80 27 2.130 1.60 63 1.845 8.60 28 2.129 1.40 64 1.868 8.40 29 2.130 1.20 65 1.920 8.20 30 2.131 1.00 66 2.009 8.00 31 2.134 0.80 67 2.208 7.80 32 2.140 0.60 68 2.531 7.60 33 2.147 0.40 69 2.915 7.40 34 2.155 0.20 70 3.299 7.20 35 2.163 0.00 71 3.683 7.00 36 2.167