Text

Cycle 20 Core Operating Limits Report
	ML15334A346
Person / Time
Site:	South Texas
Issue date:	11/12/2015
From:	Dunn R; South Texas
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NOC-AE-15003304, STI: 34235347
	Download: ML15334A346 (18)
	v • d • e

Nuclear Operating Company South Texas Pro/edt E/ectric GeneratingSLtaton PCO Box 289 Wadsworth, Texas 77483 :v/v -

November 12, 2015 NOC-AE-1 5003304 10 CFR 50.36 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 20 Core Operatincq Limits Report In accordance with Technical Specification 6.9.1 .6.d, STP Nuclear Operating Company submits the attached Core Operating Limits Report for Unit 1 Cycle 20. The report covers the core design changes made during the 1RE19 refueling outage.

There are no commitments in this letter.

Ifthere are any questions regarding this report, please contact Marilyn Kistler at (361) 972-8385 or me at (361) 972-7743.

Roland F. Dunn Manager, Nuclear Fuel &Analysis mk

Attachment:

South Texas Project Unit 1 Cycle 20 Core Operating Limits Report, Revision 0 STI: 34235347

NOC-AE-1 5003304 Page 2 of 2 CC: (electronic copy)

(paper copy)

Morgqan. Lewis & Bockius LLP Regional Administrator, Region IV Steve Frantz, Esquire U.S. Nuclear Regulatory Commission 1600 East Lamar Boulevard U.S. Nuclear Regulatory Commission Arlington, TX 76011-4511 Lisa M. Regner Lisa M. Regner NRG South Texas LP Senior Project Manager John Ragan U.S. Nuclear Regulatory Commission Chris O'Hara One White Flint North (08 H04) Jim von Suskil 11555 Rockville Pike Rockville, MD 20852 CPS Energqy Kevin Polio NRC Resident Inspector Cris Eugster U. S. Nuclear Regulatory Commission L. D. Blaylock P. 0. Box 289, Mail Code: MNIl16 Wadsworth, TX 77483 Crain Caton & James, P.C.

Peter Nemeth City of Austin Cheryl Mele John Wester Texas Dept. of State Health Services Richard A. Ratliff Robert Free

Nuclear Operating Company SOUTH TEXAS PROJECT Unit 1 Cycle 20 CORE OPERATING LIMITS REPORT Revision 0 Core Operating Limits Report Pg 1 off116 Page

l~l*Unit 1 Cycle 20 Nuclear Operating Company Core Operating Limits Report Rev. 0

1** rPage 2 of 16 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report for STPEGS Unit 1 Cycle 20 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 2.2.1 The Loop design flow for Reactor Coolant Flow-Low is 98,000 gpm.

III*

B .Unit 1 Cycle 20 Nuclear Operating Company Core Operating Limits Report Rev. 0 mr 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

-*1 measured reactor vessel AT lead/lag time constant, Ti 8 sec t2 measured reactor vessel AT lead/lag time constant, T2 --3 sec T*3 measured reactor vessel AT lag time constant, T3 = 2 sec t4 measured reactor vessel average temperature lead/lag time constant, T4 =28 sec t5s measured reactor vessel average temperature lead/lag time constant, t5 = 4 sec

-c6 measured reactor vessel average temperature lag time constant, T*6 = 2 sec K1 Overtemperature AT reactor trip setpoint, K1 = 1.14 K2 Overtemperature AT reactor trip setpoint Tavg coefficient, K2 = 0.028/°F K(3 Overtemperature AT reactor trip setpoint pressure coefficient, IK3= 0.00143/psi T' Nominal full power Tavg, T'<_ 592.0 0F 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%, fi(AD) 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 TEHERMAL 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 THERMVAL POWER. (Reference 3.6 and Section 4.4.1.2 of Reference 3.7)

Over-power AT Setpoint Parameter Values Til measured reactor vessel AT lead/lag time constant, til 8 sec T2 measured reactor vessel AT lead/lag time constant, tc2 =3 sec

-r3 measured reactor vessel AT lag time constant, T13=2 sec

-r6 measured reactor vessel average temperature lag time constant, "T6 = 2 sec

-r7 Time constant utilized in the rate-lag compensator for Tavg, "T7 --10 sec K4 Overpower AT reactor trip setpoint, IK4 = 1.08 K5 Overpower AT reactor trip setpoint Tavg rate/lag coefficient, Ks = 0.02/°F for increasing average temperature, and Ks = 0 for decreasing average temperature K6 Overpower AT reactor trip setpoint Tavg heatup coefficient K(6 = 0.002/°F for T >T", and K6 = 0 for T _ T" T" Indicated full power Tavg, T"< 592.0 0F f2 (AI) = 0 for all (AI)

N*l~TM_--* - Unit 1 Cycle 20 Nlear Operating Com.pany Core Operating Limits Report Rev. 0

1*

rPage4 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 254 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).

IDT

Unit 1 Cycle 20 Nuclear Operating Company Core Operating Limits Report Rev. 0

r 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*TP = 2.55.

2.7.2 K(Z) is provided in Figure 7.

2.7.3 The Fx limits for RATED THERMAL POWER (Fx~yP) 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 Fx 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-83 85. Therefore, these Fx 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.l11 UE = Engineering uncertainty factor of 1.03.

This uncertainty is calculated and applied automatically by the Power Distribution Monitoring System (PDMS).

IPD 3 Nuclear Operating Company Unit 1Cycle 20 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 (UJFQ) 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 F* = 1.62 2.8.2 PFAriH 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 F*H using the PDMS shall be the greater of:

UFAH= 1.04 OR UFAH = 1.0 + (UJA/I00)

Where:

UAH = Uncertainty for power peaking factor as defined in Equation 5-19 from the document referenced in Technical Specification 6.9.1.6.b.1 1.

This uncertainty is calculated and applied automatically by the Power Distribution Monitoring System.

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

4 UFAn= 1.0

N* RUnitID* 1 Cycle 20 Nuclear Operating Company Core Operating Limits Report Rev. 0

r 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 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 20 Final Reload Evaluation" NF-TG-1 5-62 (ST-UB-NOC-15003490 dated September 22, 2015.

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. 6, "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 Letter from J. M. Ralston (Westinghouse) to D. F. Hoppes (STPNOC), "South Texas Project Electric Generating Station Units 1 and 2 Documentation of the f 1(AI) Function in OTAT Setpoint Calculation," NF-TG- 11-93 (ST-UB-NOC- 11003215) dated November 10, 2011.

3.7 Document RSE-U1, Rev. 4, "Unit 1 Cycle 20 Reload Safety Evaluation and Core Operating Limits Report." (CR Action 14-10332-9)

SA 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.

3Limit 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 the QDPS display, which is bounded by the 9.6 psi averaged measurement calculated in Reference 3.4.

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

fl lUnit Nucea Oprtn Copn 1Cycee20 Core Operating Limits Report Rev. 0 Page 8 of 16 Figure 1 Reactor Core Safety Limits - Four Loops in Operation 680

,-, 620 600 0 20 40 60 80 100 120 140 Rated Thermal Power (%)

fl Nulea Oprtn lUnitl1 Cycle 20 Copn Core Operating Limits Report Rev. 0 flE D Page 9of 16 Figure 2 Required Shutdown Margin for Modes 3 & 4 7.0 6.0 5.0 C.)

3.0 20 13.0 2.0 0400 800 1200 1600 2000 2400 RCS Critical Boron Concentration (ppm)

(for ARI minus most reactive stuck rod)

fl lUnit Nucea Oprtn Copn 1Cycleo20 Core Operating Lhnits Report Rev. 0 Page 10 of 16 Figure 3 Required Shutdown Margin for Mode 5 7.0 6.0 5.0 4.0 20 1-.0 03.0 0 400 800 1200 1600 2000 2400 RCS Critical Boron Concentration (ppm)

(for ARI minus most reactive stuck rod)

- TM Unitl1Cycle 20 Nulea Oprtn Copn Core Operating Limits Report Rev. 0 Page 11 of 16 Figure 4 MTC versus Power Level 7.0 6.0 Unacceptable 5.0

"4.0 Acceptable 0

S2.0 S1.0

~0.0

-1.0

-2.0

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

IP~rBUvit1 Cycleo20 Nulea Oprtn Copn Core Operating Limits Report Rev. 0 Page 12 of 16 Figure 5 Control Rod Insertion Limits* versus Power Level 260 l l-l1 l llr* r~lli 4' I lllillllllllJl 122 Step Overlap K *(23, 259): 11 (21 ,254): 1227 ,Step Step ,Overlap Overlap~

1 - +- F - Iii +T h~

( 77,254 ): 117 Step Overlap S(79,259):

! ! [ ! I J [ !El [ I 240 V 220 200 180 160

~,140

Coto Bak i s led ihdantul u oiin O 120 Full wihrw shl:etecodto hresudw n coto bak ar at th oiino ihr24o 5 tp 0100 80 60 40 20 0 10 20 30 40 50 60 70 80 90 100 Rated Thermal Power (%)

PT Nulea Oprtn M

Copn Unitl1 Cycleo20 Core Operating Limits Report Rev. 0 Page 13 of 16 Figure 6 AFD Limits versus Power Level 120 110 (11,90) -I --

100 Uncepal Unaccptabl

- Acetal - -

90 80 70

-0 60 50 40 30 20 10 0

-50 -40 -30 -20 -10 010 20 30 40 50 Axial Flux Difference (% Delta-I)

IP~ffUnit 1Cycleo20 Nulea Oprtn Copn Core Operating Limits Report Rev. 0 Page 14 of 16 Figure 7 K(Z) - Normalized FQ (Z) versus Core Height 1.2

-- I --I I t---

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 II--~---I----t----l1 1 1 I I 1I I I 1I I I 1 I -I---

I ----

I---I-

[ -- -- I-----I-II I I I I I I I I 1.1 I I I I I I I I I I I l l I I I I I I I l l I I I : I : : I I : ; 1 1 I I I I I I I 1 1.0 -I IfIII II I II I I I -t--F I 1 I iITIII I I I II I I I I I I I[ I I I I I I I I I II I ; I ;

! I I I I* It--bI -4I- -I I I I I I I I

1 I ci 1 I I I I 0.9 I I I[ II I II I I I I I l I---I-- I--I--l I I I I I I I I I [ I I I I I I I I r I I I I I I I [ P I I I I [ I I I I 1 1 I I I r I I I I I I I I [ E I I TI I I I 0.8

[ I II I II

I I I -t--I--

I-I I---I I II I I --------I-+ ------I I I I I ;

! i!!!! 1! i ! i m ! !m ! !1 1! ] !! 1! !!

.~0.5 Core Elev. (if) FQ K(Z) 0.0 2.55 1.0 7.0 2.55 1.0

-14.0 2.359 0.925 0.4 0.3 ! 11! ! 1! ! ! ! I ! ! 1! ! I 1! I 1! ! ! 1! ! ! ! 1! ! ! ! ! ! ! 1! ! ! ! 1

-- I* I I I I I I I I I I I I II I---- I----I---I I--

I I I [ I I l I I I [II III l 0.2 ! ! 11 ! 1! ! !! I! l l L I ! ! l ! ! ! ! l ! ! ! l l ! ! ! ! ! ! 11! ! ! ! 1 0.1 I I I I : I I I I : I I I I r I I I i I ; ; ; ; I ; I ; I 1 I I ; i I ; ; I ; ; ; r I I I I I I II I I I I II 1 1 1 I I 1I I-4I1 Il1-I-- - -t -- - -- l l l 0.0 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 Core Height (ft)

JI*RUnit 1 Cycle 20 Nuclear Operating Company Core Operating Limilts Report Rev. 0

1* I *Page 15 ofl16 Table 1 (Part 1 of 2)

Unrodded Fx for Each Core Height for Cycle.Burnups Less Than 9000 MWD/MTU Core Height Axial Unrodded Core Height Axial Unrodded (Ft.) Point Fxy (Ft.) Point Fxy 14.0 1 7.3 17 6.8 37 1.972 13.8 2 5.739 6.6 38 1.993 13.6 3 4.158 6.4 39 1.975 13.4 4 2.789 6.2 40 1.932 13.2 5 2.495 6.0 41 1.90 1 13.0 6 2.226 5.8 42 1.938 12.8 7 2.138 5.6 43 1.952 12.6 8 2.118 5.4 44 1.958 12.4 9 2.068 5.2 45 2.001 12.2 10 2.022 5.0 46 2.058 12.0 11 2.002 4.8 47 2.063 11.8 12 2.014 4.6 48 2.009 11.6 13 2.037 4.4 49 1.946 11.4 14 2.014 4.2 50 1.966 11.2 15 1.965 4.0 51 1.973 11.0 16 1.933 3.8 52 1.965 10.8 17 1.925 3.6 53 1.974 10.6 18 1.920 3.4 54 2.014 10.4 19 1.918 3.2 55 2.039 10.2 20 1.93 8 3.0 56 1.99 1 10.0 21 1.973 2.8 57 1.93 8 9.8 22 1.98 1 2.6 58 1.942 9.6 23 1.943 2.4 59 1.947 9.4 24 1.908 2.2 60 1.952 9.2 25 1.904 2.0 61 1.967 9.0 26 1.895 1.8 62 1.997 8.8 27 1.896 1.6 63 2.004 8.6 28 1.916 1.4 64 1.932 8.4 29 1.983 1.2 65 1.872 8.2 30 2.036 1.0 66 1.912 8.0 31 1.977 0.8 67 2.222 7.8 32 1.925 0.6 68 3.005 7.6 33 1.929 0.4 69 4.318 7.4 34 1.945 0.2 70 6.145 7.2 35 1.947 0.0 71 9.180 7.0 36 1.939

Unt1Cycle 20 Nulea Oprtn Copn 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 9000 MWD/MTU Core Height Axial Unrodded Core Height Axial Unrodded (Ft.) Point Fxy (Ft.) Point Fxy 14.0 1 6.495 6.8 37 2.203 13.8 2 5.203 6.6 38 2.238 13.6 3 3.9 11 6.4 39 2.204 13.4 4 2.790 6.2 40 2.147 13.2 5 2.552 6.0 41 2.114 13.0 6 2.305 5.8 42 2.106 12.8 7 2.160 5.6 43 2.095 12.6 8 2.087 5.4 44 2.08 1 12.4 9 2.027 5.2 45 2.098 12.2 10 2.02 1 5.0 46 2.13 1 12.0 11 2.030 4.8 47 2.125 11.8 12 2.056 4.6 48 .2.067 11.6 13 2.082 4.4 49 2.021 11.4 14 2.074 4.2 50 2.016 11.2 15 2.044 4.0 51 2.005 11.0 16 2.009 3.8 52 1.990 10.8 17 2.03 8 3.6 53 1.992 10.6 18 2.046 3.4 54 2.026 10.4 19 2.053 3.2 55 2.048 10.2 20 2.082 3.0 56 1.990 10.0 21 2.127 2.8 57 1.935 9.8 22 2.145 2.6 58 1.905 9.6 23 2.113 2.4 59 1.875 9.4 24 2.088 2.2 60 1.864 9.2 25 2.102 2.0 61 1.880 9.0 26 2.110 1.8 62 1.926 8.8 27 2.114 1.6 63 1.954 8.6 28 2.124 1.4 64 1.932 8.4 29 2.161 1.2 65 1.945 8.2 30 2.194 1.0 66 2.054 8.0 31 2.157 0.8 67 2.418 7.8 32 2.124 0.6 68 3.143 7.6 33 2.126 0.4 69 4.250 7.4 34 2.137 0.2 70 5.782 7.2 35 2.148 0.0 71 8.469 7.0 36 2.160

Nuclear Operating Company South Texas Pro/edt E/ectric GeneratingSLtaton PCO Box 289 Wadsworth, Texas 77483 :v/v -

November 12, 2015 NOC-AE-1 5003304 10 CFR 50.36 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 20 Core Operatincq Limits Report In accordance with Technical Specification 6.9.1 .6.d, STP Nuclear Operating Company submits the attached Core Operating Limits Report for Unit 1 Cycle 20. The report covers the core design changes made during the 1RE19 refueling outage.

There are no commitments in this letter.

Ifthere are any questions regarding this report, please contact Marilyn Kistler at (361) 972-8385 or me at (361) 972-7743.

Roland F. Dunn Manager, Nuclear Fuel &Analysis mk

Attachment:

South Texas Project Unit 1 Cycle 20 Core Operating Limits Report, Revision 0 STI: 34235347

NOC-AE-1 5003304 Page 2 of 2 CC: (electronic copy)

(paper copy)

Morgqan. Lewis & Bockius LLP Regional Administrator, Region IV Steve Frantz, Esquire U.S. Nuclear Regulatory Commission 1600 East Lamar Boulevard U.S. Nuclear Regulatory Commission Arlington, TX 76011-4511 Lisa M. Regner Lisa M. Regner NRG South Texas LP Senior Project Manager John Ragan U.S. Nuclear Regulatory Commission Chris O'Hara One White Flint North (08 H04) Jim von Suskil 11555 Rockville Pike Rockville, MD 20852 CPS Energqy Kevin Polio NRC Resident Inspector Cris Eugster U. S. Nuclear Regulatory Commission L. D. Blaylock P. 0. Box 289, Mail Code: MNIl16 Wadsworth, TX 77483 Crain Caton & James, P.C.

Peter Nemeth City of Austin Cheryl Mele John Wester Texas Dept. of State Health Services Richard A. Ratliff Robert Free

Nuclear Operating Company SOUTH TEXAS PROJECT Unit 1 Cycle 20 CORE OPERATING LIMITS REPORT Revision 0 Core Operating Limits Report Pg 1 off116 Page

l~l*Unit 1 Cycle 20 Nuclear Operating Company Core Operating Limits Report Rev. 0

1** rPage 2 of 16 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report for STPEGS Unit 1 Cycle 20 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 2.2.1 The Loop design flow for Reactor Coolant Flow-Low is 98,000 gpm.

III*

B .Unit 1 Cycle 20 Nuclear Operating Company Core Operating Limits Report Rev. 0 mr 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

-*1 measured reactor vessel AT lead/lag time constant, Ti 8 sec t2 measured reactor vessel AT lead/lag time constant, T2 --3 sec T*3 measured reactor vessel AT lag time constant, T3 = 2 sec t4 measured reactor vessel average temperature lead/lag time constant, T4 =28 sec t5s measured reactor vessel average temperature lead/lag time constant, t5 = 4 sec

-c6 measured reactor vessel average temperature lag time constant, T*6 = 2 sec K1 Overtemperature AT reactor trip setpoint, K1 = 1.14 K2 Overtemperature AT reactor trip setpoint Tavg coefficient, K2 = 0.028/°F K(3 Overtemperature AT reactor trip setpoint pressure coefficient, IK3= 0.00143/psi T' Nominal full power Tavg, T'<_ 592.0 0F 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%, fi(AD) 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 TEHERMAL 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 THERMVAL POWER. (Reference 3.6 and Section 4.4.1.2 of Reference 3.7)

Over-power AT Setpoint Parameter Values Til measured reactor vessel AT lead/lag time constant, til 8 sec T2 measured reactor vessel AT lead/lag time constant, tc2 =3 sec

-r3 measured reactor vessel AT lag time constant, T13=2 sec

-r6 measured reactor vessel average temperature lag time constant, "T6 = 2 sec

-r7 Time constant utilized in the rate-lag compensator for Tavg, "T7 --10 sec K4 Overpower AT reactor trip setpoint, IK4 = 1.08 K5 Overpower AT reactor trip setpoint Tavg rate/lag coefficient, Ks = 0.02/°F for increasing average temperature, and Ks = 0 for decreasing average temperature K6 Overpower AT reactor trip setpoint Tavg heatup coefficient K(6 = 0.002/°F for T >T", and K6 = 0 for T _ T" T" Indicated full power Tavg, T"< 592.0 0F f2 (AI) = 0 for all (AI)

N*l~TM_--* - Unit 1 Cycle 20 Nlear Operating Com.pany Core Operating Limits Report Rev. 0

1*

rPage4 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 254 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).

IDT

Unit 1 Cycle 20 Nuclear Operating Company Core Operating Limits Report Rev. 0

r 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*TP = 2.55.

2.7.2 K(Z) is provided in Figure 7.

2.7.3 The Fx limits for RATED THERMAL POWER (Fx~yP) 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 Fx 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-83 85. Therefore, these Fx 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.l11 UE = Engineering uncertainty factor of 1.03.

This uncertainty is calculated and applied automatically by the Power Distribution Monitoring System (PDMS).

IPD 3 Nuclear Operating Company Unit 1Cycle 20 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 (UJFQ) 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 F* = 1.62 2.8.2 PFAriH 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 F*H using the PDMS shall be the greater of:

UFAH= 1.04 OR UFAH = 1.0 + (UJA/I00)

Where:

UAH = Uncertainty for power peaking factor as defined in Equation 5-19 from the document referenced in Technical Specification 6.9.1.6.b.1 1.

This uncertainty is calculated and applied automatically by the Power Distribution Monitoring System.

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

4 UFAn= 1.0

N* RUnitID* 1 Cycle 20 Nuclear Operating Company Core Operating Limits Report Rev. 0

r Page 7 of 16 2.9 DNB PARAMETERS (Specification 3.2.5):