NOC-AE-10002506, Cycle 16 Core Operating Limits Report, Revision 2

Cycle 16 Core Operating Limits Report, Revision 2

ML100250239
Person / Time
Site:	South Texas
Issue date:	01/19/2010
From:	Leazar D South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
32599116, NOC-AE-10002506, STI: 32599116
Download: ML100250239 (18)

Text

A'ImAFAM Nuclear Operating Company South Texas Pro/ectElectric GeneratingStation PO. Bax 289 Wadsworth, Texas 77483 -

,January 19, 2010 NOC-AE-1 0002506 10CFR50.36 STI: 32599116 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 I Cycle 16 Core Operating Limits Report In accordance with Technical Specification 6.9.1.6.d, STP Nuclear Operating Company submits the attached Revision 2 to the Core Operating Limits Report for Unit 1 Cycle

16. The report reflects a change in the control rod full out position from 259 steps to either 249 or 259 steps.

There are no commitments in this letter.

Ifthere are any questions concerning this report, please contact Marilyn Kistler at (361) 972-8385 or me at (361) 972-7795.

David A. ýaza Manager, Nuclear Fuels & Analysis mkk

Attachment:

Unit 1 Cycle 16 Core Operating Limits Report, Rev. 2 Ac o4

NOC-AE-1 0002506 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 612 East Lamar Blvd, Suite 400 Arlington, Texas 76011-8064 Mohan C. Thadani U. S. Nuclear Regulatory Commission Mohan C. Thadani Kevin Howell Senior Project Manager Catherine Callaway U.S. Nuclear Regulatory Commission Jim von Suskil One White Flint North (MS 8B1A) NRG South Texas LP 11555 Rockville Pike Rockville, MD 20852 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 C. M. Canady Jon C. Wood City of Austin Cox Smith Matthews 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

Nuclear Operating Company SOUTH TEXAS PROJECT Unit 1 Cycle 16 CORE OPERATING LIMITS REPORT Revision 2 Core Operating Limits Report Page I of 16

AlMATFM Nuclear Operating Company Unit 1Cycle 16 Core Operating Limits Report Rev. 2 Page 2 of 16 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report for STPEGS Unit 1 Cycle 16 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.

Unit 1 Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev. 2 AV f - Page 3of 16 2.2.2 The Over-temperature AT and Over-power AT. setpoint parameter values are listed below:

Over-temperature AT Setpoint Parameter Values "1 measured reactor vessel AT lead/lag time constant, c= 8 sec T2 measured reactor vessel AT lead/lag time constant, 'r2 3 see T3 measured reactor vessel AT lag time constant, 13 = 2 sec T4 measured reactor vessel average temperature lead/lag time constant, -r4 = 28 see

'r 5 measured reactor vessel average temperature lead/lag time constant, "C5 = 4 see T6 measured reactor vessel average temperature lag time constant, T6 = 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 = 0.00143/psig T' Nominal full power Tavg, T'___ 592.0 OF P' Nominal RCS pressure, P' = 2235 psig fl(A1) 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 q, 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 q% - qu 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, "l = 8 sec t2 measured reactor vessel AT lead/lag time constant, t2 = 3 see t3 measured reactor vessel AT lag time constant, "c3 = 2 sec

-r6 measured reactor vessel average temperature lag time constant, -r6 = 2 sec t7 Time constant utilized in the rate-lag compensator for Tavg, u7 = 10 see 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"_5 592.0 OF f2(AI) = 0 for all (Al)

F Unit 1 Cycle 16 Operating Company Core Operating Limits Report Rev. 2 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, IIFP, MTC shall be less negative than -62.6 pcm/IF.

2.4.3 The 300 ppm, ARO, HFP, MTC shall be less negative than -53.6 pcm/0 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/PF 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 249 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).

I t M Unit 1Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev. 2

• Page5 Nl*

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 FRTI = 2.55.

2.7.2 K(Z) is provided in Figure 7.

2.7.3 The F. limits for RATED THERMAL POWER (FRyTP) 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.

PTM Nuclear Operating Company Unit 1Cycle 16 Core Operating Limits Report Rev. 2 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):

RTP1 2.8.1 FAn = 1.62 2.8.2 PF,, = 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 + (UAH/I100)

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 1 Applies to all fuel in the Unit 1 Cycle 16 Core.

IT MB Unit 1 Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev. 2 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: I 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 16 REDESIGN Reload Evaluation (RE), Revision 1" NF-TG-09-74 Revision 1 (ST-UBNOC-09-09003012, Rev 1).

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 10-259-12, "Unit 1 Cycle 16 (Redesign) 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.

• &VA Unit 1Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev. 2 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 (%)

ýAEr" Unit 1 Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev. 2 Page 9 of 16 Figure 2 Required Shutdown Margin for Modes 3 & 4 7.0 7-6.0 AcceptableI f

/

5.0 01*1 00, 0" I I I

~4.0

/0"

~3.0 2.0

- - -T T - 1 r~-T -I T T --- r 1-I Unaccepta~bi~J f

(6010,1.30)

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

(for ARI minus most reactive stuck rod)

AIAM Unit 1Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev. 2 Page 10 of 16 Figure 3 Required Shutdown Margin for Mode 5 7.0 6.0 5.0

-s'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)

A* Unit 1 Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev. 2 O AN KPage 11 of 16 Figure 4 MTC versus Power Level 7.0.

6.0 77Unacceptable Operation 5.0 4.0

~-* lAcceptable Operation-3.0 2.0 1.0

0. -

-1.0-

-2.0 --

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

lT M Unit 1Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev. 2 Page 12 of 16 Figure 5 Control Rod Insertion Limits* versus Power Level 260 i- -I... 01( 23 "I'

, 259): 122 Step Overlap (79,259 ): 122 Step Overlap1 I -1 19 249Q 112 St Ov rlan I 1i i i li I II I -

(752494: 112 StenOverlaIn 240 I YI I 220 200 (0OA202) 180 V

160 BCoo k a B D 140 S120 100 I II 80 60 40 80 20

--- -- Fully uithdravm shall be the condition Ahere shutdoin and

--- - -- - - - control banks are at the position of either 24 9 or 259 steps (29.0) ithdrawn.

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

Unit 1 Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev. 2 Page 13 of 16 Figure 6 AN) Limits versus Power Level 120 110 p~~~i aioi 100 90 Opertio k II I I I I I 80 (-3110) -- - --

table - Unacceptable 0

70 Opertio Acceptable Operation I-.

60 50 40 30 20 10 0

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

Unit 1 Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev. 2 Page 14 of 16 Figure 7 K(Z) - Normalized FQ(Z) versus Core Height 1.2 -

1.1  !

1.0 -

0.8 a0.70.7 0.6 Core Elev.C(ft) FQ K(Z) 0.0 2.55 1.0  !!

7.0 2.55 1.0  !

• o.5 *J 14.0 2.359 0.925 0.4 ,

0.3 0 1o. 2 3 1 1 1 !3 1 0.2 0o 4 5 6 7 8 0 1 1' 13 14 Core Height (ft)

INCeainm Unit 1 Cycle 16 Nuclea Operating Company Core Operating Limits Report Rev. 2 Page 15 of16 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 6.759 6.80 37 1.916 13.80 2 5.394 6.60 38 1.911 13.60 3 4.029 6.40 39 1.907 13.40 4 2.663 6.20 40 1.904 13.20 5 2.079 6.00 41 1.902 13.00 6 1.839 5.80 42 1.902 12.80 7 1.860 5.60 43 1.903 12.60 8 1.907 5.40 44 1.907 12.40 9 1.935 5.20 45 1.913 12.20 10 1.921 5.00 46 1.920 12.00 11 1.907 4.80 47 1.927 11.80 12 1.902 4.60 48 1.930 11.60 13 1.911 4.40 49 1.931 11.40 14 1.918 4.20 50 1.933 11.20 15 1.919 4.00 51 1.936 11.00 16 1.913 3.80 52 1.937 10.80 17 1.910 3.60 53 1.935 10.60 18 1.910 3.40 54 1.932 10.40 19 1.913 3.20 55 1.935 10.20 20 1.922 3.00 56 1.941 10.00 21 1.940 2.80 57 1.944 9.80 22 1.962 2.60 58 1.944 9.60 23 1.982 2.40 59 1.936 9.40 24 1.995 2.20 60 1.919 9.20 25 2.005 2.00 61 1.866 9.00 26 2.015 1.80 62 1.803 8.80 27 2.025 1.60 63 1.736 8.60 28 2.036 1.40 64 1.735 8.40 29 2.048 1.20 65 1.743 8.20 30 2.059 1.00 66 1.774 8.00 31 2.054 0.80 67 1.908 7.80 32 2.034 0.60 68 2.120 7.60 33 1.995 0.40 69 2.370 7.40 34 1.957 0.20 70 2.621 7.20 35 1.940 0.00 71 2.871 7.00 36 1.925

• 1T! Unit 1 Cycle 16 Nuclear Operating Company Core Operating Limits Report Rev. 2 N* AF rPage 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 5.581 6.80 37 2.050 13.80 2 4.739 6.60 38 2.052 13.60 3 3.897 6.40 39 2.047 13.40 4 3.055 6.20 40 2.039 13.20 5 2.553 6.00 41 2.029 13.00 6 2.247 5.80 42 2.016 12.80 7 2.147 5.60 43 2.004 12.60 8 2.105 5.40 44 1.992 12.40 9 2.073 5.20 45 1.982 12.20 10 2.024 5.00 46 1.972 12.00 11 1.999 4.80 47 1.963 11.80 12 1.981 4.60 48 1.953 11.60 13 1.968 4.40 49 1.944 11.40 14 1.964 4.20 50 1.935 11.20 15 1.960 4.00 51 1.927 11.00 16 1.958 3.80 52 1.920 10.80 17 1.957 3.60 53 1.914 10.60 18 1.955 3.40 54 1.909 10.40 19 1.956 3.20 55 1.901 10.20 20 1.968 3.00 56 1.891 10.00 21 1.987 2.80 57 1.879 9.80 22 2.009 2.60 58 1.863 9.60 23 2.024 2.40 59 1.847 9.40 24 2.029 2.20 60 1.829 9.20 25 2.030 2.00 61 1.813 9.00 26 2.027 1.80 62 1.798 8.80 27 2.025 1.60 63 1.788 8.60 28 2.022 1.40 64 1.823 8.40 29 2.019 1.20 65 1.849 8.20 30 2.017 1.00 66 1.903 8.00 31 2.016 0.80 67 2.160 7.80 32 2.016 0.60 68 2.609 7.60 33 2.022 0.40 69 3.153 7.40 34 2.029 0.20 70 3.696 7.20 35 2.038 0.00 71. 4.240 7.00 36 2.045