NOC-AE-08002360, Cycle 14 Core Operating Limits Report

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Cycle 14 Core Operating Limits Report
ML083170486
Person / Time
Site: South Texas STP Nuclear Operating Company icon.png
Issue date: 11/04/2008
From: Leazar D
South Texas
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-08002360
Download: ML083170486 (18)
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Nuclear Operating Company South Texas Project Electric GeneratinS Station P.. Bax 289 Wadsworth, T7,as 77483 -

November 4, 2008 NOC-AE-08002360 10CFR50.36 STI: 32388355 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 14 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 14. This report reflects core design changes made during the 2RE13 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. Leazar Manager, Nuclear Fuels & Analysis mk

Attachment:

Unit 2 Cycle 14 Core Operating Limits Report, Rev. 0 f'.

NOC-AE-08002360 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 76011-4125 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 7 D1) NRG South Texas LP 11555 Rockville Pike Rockville, MD 20852 Ed Alarcon Senior Resident Inspector J. J. Nesrsta U. S. Nuclear Regulatory Commission R. K. Temple P. 0. Box 289, Mail Code: MN1 16 Kevin Polio Wadsworth, TX 77483 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

Nuclear Operating Company SOUTH TEXAS PROJECT Unit 2 Cycle 14 CORE OPERATING LIMITS REPORT Revision 0 Core Operating Limits Report Page I of 16

SP'rAN Unit 2 Cycle 14 MENI-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 14 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.

Nla OUnit 2 Cycle 14 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 measured reactor vessel AT lead/lag time constant, 1u1 = 8 sec measured reactor vessel AT lead/lag time constant, T2 = 3 sec

ýT3 measured reactor vessel AT lag time constant, "T3= 2 sec measured reactor vessel average temperature lead/lag time constant, "r4 = 28 sec T5 measured reactor vessel average temperature lead/lag time constant, Irs = 4 sec measured reactor vessel average temperature lag time constant, Tr 6 = 2 sec K' Overtemperature AT reactor trip setpoint, K, = 1.14 K(2 Overtemperature AT reactor trip setpoint Tavg coefficient, K2 = 0.028/7F K(3 Overtemperature AT reactor trip setpoint pressure coefficient, K3 = 0.001 43/psig Nominal full power Tavg, T'< 592.0 'F Nominal RCS pressure, P' = 2235 psig f,(AD) 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 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 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, ij = 8 sec Tr2 measured reactor vessel AT lead/lag time constant, 1r2 = 3 sec T3 measured reactor vessel AT lag time constant, Tr3 = 2 sec T6 measured reactor vessel average temperature lag time constant, '16 = 2 sec T7 Time constant utilized in the rate-lag compensator for Tavg, '17 = 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 = 0.002/0 F for T>T",andK6 = 0forT_ T" T" Indicated full power Tavg, T"< 592.0 'F f 2(AI) = 0 for all (AD)

drifFAN Unit 2 Cycle 14 Nuclear Operating Company 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 l'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 pcmIF.

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 Qut, 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 ppmMTC 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 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).

dLW1 3A Nuclear Operating Company Unit 2 Cycle 14 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):

RTP 2.7.1 FQ =2.55.

2.7.2 K(Z) is provided in Figure 7.

2.7.3 The Fxy limits for RATED THERMAL POWER (FxyRTP 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.

APDrAM Unit 2 Cycle 14 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):

RTP 2.8.1 FAH = 1.62 2.8.2 PF,, H 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 (UFrH) to be applied to the FA using the PDMS shall be calculated by:

UFAH = 1.0 + (UAH/IO0)

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 Applies to all fuel in the Unit 2 Cycle 14. Core.

A lAFiMb Unit 2 Cycle 14 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 -* 595 °F 2 2.9.1.2 Pressurizer Pressure > 2200 psig ,

2.9.1.3 Minimum Measured Reactor Coolant System Flow > 403,000 gpm 4 .

3.0 REFERENCES

3.1 Letter from D. V. Lockridge (Westinghouse) to D. F. Hoppes (STPNOC), "South Texas Project Nuclear Operating Company South Texas Project Electric Generating Station Unit 2 Cycle 14 Final Reload Evaluation, " ST-UB-NOC-08002914, October 14, 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.

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 07-8330-9, Revision 0, "Unit 2 Cycle 14 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.

3.7 5Z529ZB01025, 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 OF 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 per Reference 3.7.

Unit 2 Cycle 14 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 (%)

P *rA

  • Unit 2 Cycle 14 Nucear Operating Company "Core Operating Limits Report Rev. 0 Page 9 of 16 Figure 2 Required Shutdown Margin for Modes 3 & 4 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)

N.WAIF O M CUnit 2 Cycle 14 Nuclear Operating Company Core Operating Li"its 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 I-2.0 1.0 0.0 0 W- 400 ) 800 1200 1600 2000 2400 RCS Critical Boron Concentration (ppm)

(for ARI minus most reactive stuck rod)

ldr~ llM Unit 2 Cycle 14 Nuclear Operating Company Core Operating Limits Report Rev. 0 A N Page 11 of 16 Figure 4 MTC versus Power Level 7.0 6.0 6.0Unacceptable Operationi 5.0 4.0-0 lAcceptable Operationj 3.0-2.0 1.0 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 14 Nuclear Operating Company Core Operating Limits Report Rev. 0 MENf f- Page 12 of 16 Figure 5 Control Rod Insertion Limits* versus Power Level 260 r I. .. IL I .. . . .

  • L , i k +--- 7 01 ( 23 , 259 ): 122 Step Overlap ( 79,259): 122 Step Overlap F

OF: I l lZ 1 1 1 1 1 1

240 220 200 (0, 202)11, 2i :

[Jo,

[ I J I ~ l I OF 180 OFl (100 174 00e 0I I

- -:  :-] an C i-2

"' 160 1000

.2 140 i[ l l

  • iI l 1 "00"00Y I II I I l l

= 120 100 80 OFlI I  :

60 0, 1[

[ IOl I 40

' , OF ontrol Bank A is alredy with rw to FuH Out Position.

20 .... 00 Fully withdrawn shall be the condition where shutdown and FT- -- -- -- -- contrl banks are at the position of 259 steps withrw.

0~

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

~dL3 Unit 2Cycle 14 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 13 of 16 Figure 6 AFD Limits versus Power Level 120 I I I k 110 100

( -11,90-90 80 Unacceptable Unacceptable Operation Aceeptable Operation 70 00 Operation ----

60

(-1, (31 50 50 40 30 20 10 0

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

Unit 2 Cycle 14 Nl r O y Core Operating Limits Report Rev. 0 Page 14 of 16 Figure 7 K(Z) - Normalized FQ (Z) versus Core Height 1.2 II L I 1-1-1-1 J I I 111 11 11 1 1 1 1_LL -. I I LLLL

--- ITLU 7TTLL TFTI 1.1 f i ! i i -1 i i i i i i i i ! i i i ! ! i i ! ! ! i ! i i i i ! i ! i i i i iA i i i i i i i i ! i i ! i-i ! i i ! i i i ! i i i i i I 1.0 i i i i i i I i ! iI i i i i I  ! Ij i i=F E ,I= ! i Ii , i I i iI ,I I I i I 0.9 i i i i iIi i i !I i  ! Ii i i i i  ! ! !IIi i I I

!  !!i I !  ! ! II I !I ii

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.I III II II I I I . . . I .. .. II I. II I .

.I II I . .I I. . Ii II. l.I I I I I I II

'0 60.

Core Elev. (ft) FQ K(Z),

0.0 2.55 1.0 7.0 2.55 1.0 14.0 2.359 0.925 0.5 0.4.

0.3 ii i! I !I -i i i ii i i i Ii!

i !I i i ii iI i l i  ; i i iii i  ! IIi i IIiii ii i i i i ! i ! ii ! ! I.

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!_  ! i ; i i!  ! i i !I i  ! !

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!_ ! 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 i i i i i i i ! i i i i i j  ! i  ! i i i I I I I I I I I I I I I I I I I I I I I II II III II I I I I I I II IIII I I I I I 0.0 0 1 2 3 4 5 6 7' 8 9 .10 11 12 13 14 Core Height (ft)

Unit 2 Cycle 14 Nuclear Operating Company Core Operating Limits Report Rev. 0 Pg 1--

Page 15 of 16 Table 1 (Part 1 of 2)

Unrodded F.y 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 5.679 6.80 37 2.001 13.80 2 4.741 6.60 38 1.996 13.60 3 3.803 6.40 39 1.994 13.40 4 2.865 6.20 40 1.994 13.20 5 2.423 6.00 41 1.992 13.00 6 2.201 5.80 42 1.990 12.80 7 2.136 5.60 43 1.992 12.60 8 2.104 5.40 44 1.995 12.40 9 2.077 5.20 45 2.001 12.20 10 2.049 5.00 46 2.003 12.00 11 2.019 4.80 47 2.003 11.80 12 1.998 4.60 48 2.000 11.60 13 1.987 4.40 49 1.996 11.40 14 1.976 4.20 50 1.995 11.20 15 1.962 4.00 51 1.994 11.00 .16 1.945 3.80 52 1.990 10.80 17 1.934 3.60 53 1.982 10.60 18 1.928 3.40. 54 1.969 10.40 19 1.925 3.20 55 1.960 10.20 20 1.930 3.00 5.6 1.952 10.00 21 1.944 2.80 57 1.946 9.80 22 1.964 2.60 58 1.931 9.60 23 1.982 2.40 59 1.907 9.40 24 1.996 2.20 60 1.875 9.20 25 2.008 2.00 61 1.828 9.00 26 2.020 1.80 62 1.778 8.80 27 2.025 1.60 63 1.731 8.60 .28 2.029 1.40 64 1.728 8.40 29 2.044 1.20 65 1.728 8.20 30 2.058 1.00 66 ,1.746 8.00 31 2.061 0.80 67 1.867 7.80 32 2.053 0.60 68 2.076 7.60 33 2.039 0.40 69 2.329 7.40 34 2.028 0.20 70 2.581 7.20 35 2.015 0.00 71 .2.834 7.00 36 2.006

AP AIFA Unit 2 Cycle 14 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 MWDJMTU Core Height Axial Unrodded Core Height Axial Unrodded (Ft.) Point Fxy (Ft.) Point Fxy 14.00 1 4.234 6.80 37 2.170 13.80 2 3.776 6.60 38 2.165 13.60 3 3.317 6.40 39 2.150 13.40 4 2.859 6.20 40 2.135 13.20 5 2.550 6.00 41 2.121 13.00 6 2.328 5.80 42 2.109 12.80 7 2.202 5.60 43 2.098 12.60 8 2.124 5.40 44 2.088 12.40 9 2.063 5.20 45 2.077 12.20 10 2.022 5.00 46 2.066 12.00 11 1.989 4.80 47 2.054 11.80 12 1.988 4.60 48 2.042 11.60 13 1.988 4.40 49 2.031 11.40 14 1.992 4.20 50 2.020 11.20 15 1.994 4.00 51 2.010 11.00 16. 1.998 3.80 52 1.998 10.80 17 2.001 3.60 53 1.986 10.60 18 2.006 3.40 54 1.973 10.40 19, 2.011 3.20 55 1.961 10.20 20 2.026 3.00 56 1.945 10.00 21 2.046 2.80 57 1.928 9.80 22 2.067 2.60 58 1.923 9.60 23 2.084 2.40 59 1.902 9.40 24 2.094 2.20 60 1.877 9.20 25 2.101 2.00 61 1.868 9.00 26 2.106 1.80 62 1.865 8.80 27 2.111 1.60 63 1.867 8.60 128 2.117 1.40 64 1.882 8.40 29 2.122 1.20 65 1.932 8.20 30 2.127, 1.00 66 2.019 8.00 31 2.133 0.80 67 2.201 7.80 -32 2.139 0.60 68 2.488 7.60 33 2.147 0.40 69 2.827 7.40 34 2.156 0.20 70 3.166 7.20- 35 2.165 0.00 71 3.506 7.00 36 2.170

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