NOC-AE-12002936, Submittal of Cycle 18 Core Operating Limits Report

Submittal of Cycle 18 Core Operating Limits Report

ML12341A231
Person / Time
Site:	South Texas
Issue date:	11/28/2012
From:	Dunn R South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-12002936, STI: 33632179
Download: ML12341A231 (18)

Text

Nuclear Operating Company South Texas Prolect Electric GeneratingStation PO.Ba' 289 Wadsworth, Texas 77483 -

November 28, 2012 NOC-AE-12002936 STI: 33632179 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 South Texas Project Unit 1 Docket No. STN 50-498 Unit 1 Cycle 18 Core Operatinq 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 18. The report covers the core design changes made during the 1RE17 refueling outage.

There are no commitments included in this letter.

If there are any questions on this report, please contact either Marilyn Kistler at (361) 972-8385 or me at (361) 972-7743.

Roland F. Dunn Manager, Nuclear Fuel & Analysis MK

Attachment:

Revision 0 Unit 1 Cycle 18 Core Operating Limits Report

NOC-AE-12002936 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 Balwant K. Singal U. S. Nuclear Regulatory Commission Balwant K. Singal John Ragan Senior Project Manager Chris O'Hara U.S. Nuclear Regulatory Commission Jim von Suskil One White Flint North (MS 8 B1) NRG South Texas LP 11555 Rockville Pike Rockville, MD 20852 Kevin Polio Senior Resident Inspector Richard Pena U. S. Nuclear Regulatory Commission City Public Service P. 0. Box 289, Mail Code: MN1 16 Wadsworth, TX 77483 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 l

MA M=

NA-Nuclear Operating Company SOUTH TEXAS PROJECT Unit 1 Cycle 18 CORE OPERATING LIMITS REPORT Revision 0 Core Operating Limits Report Page I of 16

WOVENR Unit 1 Cycle 18 Nuclear Operating Company Core Operating Limits Report Rev. 0

ý N ffr Page 2 of 16 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report for STPEGS Unit 1 Cycle 18 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.

T Mll*l Nuclear Operating Company Unit 1 Cycle 18 Core Operating Limits Report Rev. 0 AF r 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 TI measured reactor vessel AT lead/lag time constant, cl = 8 sec T2 measured reactor vessel AT lead/lag time constant, '2 = 3 sec T3 measured reactor vessel AT lag time constant, T3 = 2 sec T4 measured reactor vessel average temperature lead/lag time constant, T4 = 28 sec T5 measured reactor vessel average temperature lead/lag time constant, r5 = 4 sec 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, K 3 = 0.00143/psig T' Nominal full power Tavg, T'_* 592.0 OF P' Nominal RCS pressure, P' = 2235 psig fi(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%, 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,- 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, ul = 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, 'C6 = 2 sec T7 Time constant utilized in the rate-lag compensator for Tavg, T7 = 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 OF f 2(AI) = 0 for all (AI)

VOWIAM Unit 1Cycle 18 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 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/IF.

2.4.3 The 300 ppm, ARO, HFP, MTC shall be less negative than -53.6 pcm/IF (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 258 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).

WOW E Nuclear Operating Company Unit I Cycle 18 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 F' -2.55.

2.7.2 K(Z) is provided in Figure 7.

2.7.3 The Fxy limits for RATED THERMAL POWER (Fxy) 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 Fy 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/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.

ApI*r~ Unit 1 Cycle 18 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):

2.8.1 FT 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 the greater of:

UF~A = 1.04 OR UFAH = 1.0 + (UA-/100)

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 thepower distribution monitoring system.

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

UFrA= 1.04 Applies to all fuel in the Unit 1 Cycle 18 Core.

lp lr~m Unit 1Cycle 18 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 OF 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. M. Ralston (Westinghouse) to D. F. Hoppes (STPNOC), "South Texas Project Electric Generating Station Unit 1 Cycle 18 Final Reload Evaluation" NF-TG-12-68 (ST-UB-NOC-12003275) dated August 13, 2012.

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 5Z529ZB0 1025 Rev. 4, Design Basis Document, Technical Specifications /LCO, Tech Spec Section 3.2.5.c 3.6 Document RSE-U1, Rev.0, "Unit I Cycle 18 Reload Safety Evaluation and Core Operating Limits Report", CR Action 11-6629-9 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, Page 37.

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. Per Technical Specification 3.2.5 Bases this includes a 10.7 psi measurement uncertainty as read on QDPS display, which is bounded by the 9.6 psi averaged measurement uncertainty calculated in Reference 3.4.

4 Includes the most limiting flow measurement uncertainty of 2.8% from Reference 3.5.

Unit 1 Cycle 18 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 8 of 16 Figure 1 Reactor Core Safety Limits - Four Loops in Operation 680

-. 620 0

S600 0 20 40 60 80 100 120 140 Rated Thermal Power (%)

Na i C Unit 1 Cycle 18 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 9 of 16 Figure 2 Required Shutdown Margin for Modes 3 & 4 7.0 4.0 3.0 S

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

(for ARI minus most reactive stuck rod)

9PIIrA Unit I Cycle 18 Nuclear Operating Company Core Operating Limits Report Rev. 0 SAF Nr Page 10 of16 Figure 3 Required Shutdown Margin for Mode 5 0 400 800 1200 1600 2000 2400 RCS Critical Boron Concentration (ppm)

(for ARI minus most reactive stuck rod)

TlVAM Unit 1Cycle 18 Nuclear Operating Company Core Operating Limits Report Rev. 0

ý A ffPage 11 of 16 Figure 4 MTC versus Power Level 7.0 6.0 Unacceptable 5.0 \

4.0

~IAcce table 0

• 2.0 1.0-0.0o

-1.0

-2.0

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

A lV AP Unit 1Cycle 18 Nuclear Operating Company Core Operating Limits Report Rev. 0

ý AV f - Page 12 of16 Figure 5 Control Rod Insertion Limits* versus Power Level 260 23 ,259): 122 Step Overlap I II ,I~,- (79,259): 122 Step Overlap

- --l----+---+ - I t -+/---I-~

( 22,258 ): 121 Step Overlap 78 258): 121 Step Overlap I

240 I 1i 11041 1 I 220 Bank B oezzzziz 200 180 - 100,174)1

• 160 BankC--------

. 140

= 120 _ _e 1 0 11 11 100 80 60 0,65) 40

.4- Control Bank Ais already withdrawn to Full Out Position.

20 -- - - - - - - - - - - Fully withdrawn shall be the condition where shutdown and

--- - - - - - - - -- control banks are at the position of either 258 or 259 steps I0thdrawn.

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

A&WE Unit I Cycle 18 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 13 of 16 Figure 6 AFD Limits versus Power Level 120 110 100

- - (-19)--- - - 1141,9o) _

90 80 Unacceptable- Unacceptable

" 70 IIII~IIIAcceptable 1~ IF I ,

S60 50 1 J1i l I I .

40 30 20 10 0

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

,C Unit 1 Cycle 18 Nuclear Operaing Company Core Operating Limits Report Rev. 0 Page 14 of 16 Figure 7 K(Z) - Normalized FQ(Z) versus Core Height 1.2 U = mi i i i Ri i i i-1= 11 I 1 11 il 1111 - I' Jill 1111.1 1.1 1 1 M I 111 1 1 1 1 1 1 11 1 1' 1l 1 1 '1 J' 11 I 11111111 I 1' IIMI 1.0 I I I I I I I I I 0.9 j i i i i i! i I i i i i i !i i i i ! i i i i1 1 1 1 . . . . . . . . . . . . . . . . . . . . . I 0.8

-I I LLII III 0.7 0.6 Core Elev.(ft) FQ K(Z) 0.0 2.55 1.0 7.0 2.55 1.0

.i 0.5 14.0 2.359 0.925 0.4 0.3 :1 ci I C :1 0.2 0.1 0.0 iit 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 Core Height (ft)

WnDr3 Unit 1Cycle 18 Nuclear Operating Company Core Operating Limits Report Rev. 0 l N Ar- 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 Fxv I4-(Ft.) Point Fxy 14.00 1 6.380 6.80 37 1.898 13.80 2 5.293 6.60 38 1.894 13.60 3 4.207 6.40 39 1.892 13.40 4 3.121 6.20 40 1.892 13.20 5 2.587 6.00 41 1.893 13.00 6 2.225 5.80 42 1.895 12.80 7 2.183 5.60 43 1.895 12.60 8 2.111 5.40 44 1.896 12.40 9 2.054 5.20 45 1.894 12.20 10 2.007 5.00 46 1.892 12.00 11 1.972 4.80 47 1.892 11.80 12 1.951 4.60 48 1.893 11.60 13 1.947 4.40 49 1.897 11.40 14 1.940 4.20 50 1.899 11.20 15 1.928 4.00 51 1.905 11.00 16 1.910 3.80 52 1.912 10.80 17 1.900 3.60 53 1.919 10.60 18 1.895 3.40 54 1.923 10.40 19 1.893 3.20 55 1.925 10.20 20 1.901 3.00 56 1.927 10.00 21 1.918 2.80 57 1.931 9.80 22 1.940 2.60 58 1.939 9.60 23 1.955 2.40 59 1.943 9.40 24 1.965 2.20 60 1.945 9.20 25 1.959 2.00 61 1.942 9.00 26 1.951 1.80 62 1.928 8.80 27 1.943 1.60 63 1.906 8.60 28 1.937 1.40 64 1.893 8.40 29 1.932 1.20 65 1.899 8.20 30 1.931 1.00 66 1.938 8.00 31 1.930 0.80 67 2.102 7.80 32 1.931 0.60 68 2.544 7.60 33 1.930 0.40 69 3.124 7.40 34 1.926 0.20 70 3.704 7.20 35 1.915 0.00 71 4.284 7.00 36 1.904

tU Ppn nit 1 Cycle 18 Nar erat n 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 Fxv (Ft.) Point Fxy -

I 14.00 1 6.496 6.80 37 2.132 13.80 2 5.468 6.60 38 2.130 13.60 3 4.441 6.40 39 2.131 13.40 4 3.413 6.20 40 2.130 13.20 5 2.770 6.00 41 2.127 13.00 6 2.247 5.80 42 2.118 12.80 7 2.161 5.60 43 2.107 12.60 8 2.076 5.40 44 2.096 12.40 9 2.027 5.20 45 2.085 12.20 10 1.990 5.00 46 2.075 12.00 11 1.977 4.80 47 2.065 11.80 12 1.976 4.60 48 2.055 11.60 13 1.981 4.40 49 2.042 11.40 14 1.990 4.20 50 2.028 11.20 15 1.999 4.00 51 2.013 11.00 16 2.010 3.80 52 1.999 10.80 17 2.018 3.60 53 1.986 10.60 18 2.024 3.40 54 1.975 10.40 19 2.029 3.20 55 1.961 10.20 20 2.042 3.00 56 1.945 10.00 21 2.060 2.80 57 1.925 9.80 22 2.081 2.60 58 1.898 9.60 23 2.098 2.40 59 1.869 9.40 24 2.111 2.20 60 1.840 9.20 25 2.120 2.00 61 1.825 9.00 26 2.125 1.80 62 1.815 8.80 27 2.126 1.60 63 1.810 8.60 28 2.125 1.40 64 1.830 8.40 29 2.124 1.20 65 1.833 8.20 30 2.124 1.00 66 1.863 8.00 31 2.125 0.80 67 2.140 7.80 32 2.128 0.60 68 2.662 7.60 33 2.136 0.40 69 3.305 7.40 34 2.141 0.20 70 3.948 7.20 35 2.142 0.00 71 4.591 7.00 36 2.136