NOC-AE-14003124, Unit 1 Cycle 19 Core Operating Limits Report

Unit 1 Cycle 19 Core Operating Limits Report

ML14127A010
Person / Time
Site:	South Texas
Issue date:	04/23/2014
From:	Dunn R South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-14003124
Download: ML14127A010 (18)

Text

Nuclear Operating Company South Texas Pro,'ect Electric GeneratinS Station PC. 7o 28,9 Wadsworth, Texas 77483 VVVV April 23, 2014 NOC-AE-14003124 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 19 Core Operating 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 19. The report covers the core design changes made during the 1RE18 refueling outage.

There are no commitments included in this letter.

If there are any questions on this report, please contact Wendy Brost at (361) 972-8516 or me at (361) 972-7743.

Roland F. Dunn Manager, Nuclear Fuel & Analysis web

Attachment:

Unit 1 Cycle 19 Core OperatingLimits Report, Revision 0 A00 (

ýJm STI: 33857840

NOC-AE-1 4003124 Page 2 of 2 cc: (electronic copy)

(paper copy)

Regional Administrator, Region IV A. H. Gutterman, Esquire U. S. Nuclear Regulatory Commission Morgan, Lewis & Bockius LLP 1600 East Lamar Boulevard Arlington, TX 76011-4511 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 NRC Resident Inspector Kevin Polio U. S. Nuclear Regulatory Commission Cris Eugster P. O. Box 289, Mail Code: MN116 L. D. Blaylock Wadsworth, TX 77483 City Public Service Jim Collins 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 Robert Free Texas Department of State Health Services

Nuclear Operating Company SOUTH TEXAS PROJECT Unit 1 Cycle 19 CORE OPERATING LIMITS REPORT Revision 0 Core Operating Limits Report Page I of 16

[p ir" lUnit 1Cycle 19 Nuclear Operating Company Core Operating Linits Report Rev. 0 APage 2 of 16 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report for STPEGS Unit 1 Cycle 19 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.

dp ir"Unit Nuclear Operating Company 1 Cycle 19 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 T1 measured reactor vessel AT lead/lag time constant, T1 = 8 sec 3T2 measured reactor vessel AT lead/lag time constant, 'r2 = 3 sec T3 measured reactor vessel AT lag time constant, 3-3= 2 sec T4 measured reactor vessel average temperature lead/lag time constant, -34 = 28 sec 3;5 measured reactor vessel average temperature lead/lag time constant, T5 = 4 sec T6 measured reactor vessel average temperature lag time constant, 3;6 = 2 sec K, 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/psi T' Nominal full power Tavg, T'< 592.0 OF P' Nominal RCS pressure, P' = 2235 psig f,(Al) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chamnbers; 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 q,+ qb is total THERMAL POWER in percent of RATED THERMAL POWER; (2) For each percent that the magnitude of q,- 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. (Reference 3.6 and Section 4.4.1.2 of Reference 3.7)

Over-power AT Setpoint Parameter Values TI1 measured reactor vessel AT lead/lag time constant, T1, = 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

'r6 measured reactor vessel average temperature lag time constant, 'r6 = 2 sec

'17 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",and K6 = 0forT_< T" T" Indicated full power Tavg, T"< 592.0 OF f2(AI) = 0 for all (Al)

Unit 1 Cycle 19 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/°F.

2.4.3 The 300 ppm, ARO, HFP, MTC shall be less negative than -53.6 pcm/OF (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 algorithmn from the document referenced by Technical Specification 6.9.1.6.b. 10:

Revised Predicted MTC = Predicted MTC + AFD Correction - 3 pcln/°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).

ApIDII Unit 1 Cycle 19 Nuclear Operating Company 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 detennined 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):

RIP 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 ( )Rwithin 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 Fy 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 firom the document referenced by Technical Specification 6.9.1.6.b. 1I UE = Engineering uncertainty factor of 1.03.

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

• llll* Unit 1 Cycle 19 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 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:

UFAH = 1.04 OR UFAH = 1.0 + (UAH/100)

Where:

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

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:

UFAH = 1.04

S W"i Nuclear Operating Company Unit 1 Cycle 19 Core Operating Linits 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 0F 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 19 Final Reload Evaluation" NF-TG-14-21 (ST-UB-NOC-14003394 dated February 20, 2014.

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. 5, "Loop Uncertainty Calculation for Narrow Range Pressurizer Pressure Monitoring Instrumentation," Section 2.3, Page 9.

3.5 5Z529ZB01025 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, (AI) Function in OTAT Setpoint Calculation," NF-TG- 11-93 (ST-UB-NOC- 11003215) dated November 10, 2011.

3.7 Document RSE-U1, Rev. 2, "Unit 1 Cycle 19 Reload Safety Evaluation and Core Operating Limits Report." (CR Action 13-531 -9)

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

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

• p ill[ Unit 1 Cycle 19 Nuclear Operating Company Core Operating Limits Report Rev. 0 W- PN-6 Pa~ge 8 of 16 Figure 1 Reactor Core Safety Limits - Four Loops in Operation 680 0

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

dr A Unit 1 Cycle 19 Nuclear 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 I-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)

d lWi l Unit 1 Cycle 19 Nuclear Operating Company Core Operating Lin-its Report Rev. 0 Page 10 of 16 Figure 3 Required Shutdown Margin for Mode 5 7.0 6.0 5.0 4.0 S3.0 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)

WlIMM Unit 1Cycle 19 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 1Iof 16 Figure 4 MTC versus PowerLevel 7.0 6.0 Unacceptable 5.0 "4.0 Acceptable

• 3.0

- 2.0

• 1.0 I0.0

-1.0

-2.0 4- ~ I - + 4- t- - I- +- 4- + - - f-

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

Arlp!M Unit 1 Cycle 19 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 12 of 16 Figure 5 Control Rod Insertion Limits* versus Power Level 260 I l l I lI l l lIlI .I ..........

- ..I I I I l l l I l l l l lA (79,259):..122 SteplOverlap I I I I - I I - I 00 -(79, 259 ):122 Step Overlap (21,254): g l -( 77,254): 117 Step Overlap 259): 117StepOverlap 1(23, 122 Step Overlap I I l l l l I I l l*

240 A Of f iF[1 fl 00, 220 wO Bank B OF or 200 M 202 1,00, 1 1 180 100, 1-74)

-T-F-F-F-T-

"0F 160 O-e C

--- - - - ooe--

100 I Bank I c, 140

,0 O120 "00, 100 oe oe

-4+

80 ?0 iamn 4B 00, 00, 60 1 '00, 0,65) 1 40 80 OF or =M . Alt 20 A I* Control Bank A is already withdrawn to Full Out Position.

and the condition where shutdown

- Fully withdrawn shallthebeposition of either 254 or 259 steps (29,0) - control banks are at IM - ý00 withdrawn.

-H+

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

Unit 1 Cycle 19 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 13 of 16 Figure 6 AFD Limits versus Power Level 120 I I I I I I i i i i i i i i i i i-110 .

I

. . . . . . . . . . I I I I I I I 100

(-1 $0 11 90 90 80 Unacceptable Unacceptable 1~ 70 - - Acceptable 0

E 60 50 40 30

.I I ET 20 10 0

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

A lPAlI 3A Unit 1Cycle 19 Nuclear Operating Company Core Operating Limits Report Rev. 0 i Ai N-- -Page 14 of 16 Figure 7 K(Z) - Normalized FQ(Z) versus Core Height 1.2 1.1 ............i i i i i i i ! . . .

i i i ! i ! i I I ii ý I i i!  ! ! i ! ! i I i ! i i 1.0

! II I! ! ! ! 1 1 II 1 1 1- - - - - 1 1 ! f 4 4 I---! 1 1 - !- 4-*

1 --- 4 -1* - - J' 1 1 -1 I- I - . f . -. I I! iii i- ii i H !I I i I i h WhI ii ii if i ! i  ! i ! i iI 0.9 i: :1 , I: :1 , :1I i. , , , , 1.. . . .. . . . .

0.8 I  ::::::::I:::::::::I :I::::I.:: I  ::I::::I::::,:. I I  :::I I

6. 0.7 ...........................................................  ::.:I::,':................

=0.6 - ore Elev. (ft) FQ K(Z)

__ _0.0 2.55 1.0 7.0 2.55 1.0 - - - -

.t 0.5

_- _ _14.0 2.359 0.925 z 0.4 0.3 rn-mi rnmmnmmI .... I Tvwrvnvl 0.2 0 . I I I M, 1 0 1 1 0 1 1 1 MI I 1 0 , I , 0 , ,  ; Ii I I I 1011i 10111; 0; ; I I i I i . i 0i

ý=r m=I~h tnhlI ~~ m 0.1 0.0 14 H +1 1 H i i i i i I 4 4 T T ,4 + H 4ii ii 1 1 H 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 Core Height (ft)

Nuclear l IM M Operating Company Unit 1Cycle 19 Core Operating Limits Report Rev. 0 Page 15 of 16 Table I (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 Fxy (Ft.) Point Fxv 14.0 1 7.641 6.8 37 1.870 13.8 2 5.953 6.6 38 1.919 13.6 3 4.255 6.4 39 1.901 13.4 4 2.796 6.2 40 1.855 13.2 5 2.535 6.0 41 1.842 13.0 6 2.331 5.8 42 1.855 12.8 7 2.208 5.6 43 1.859 12.6 8 2.123 5.4 44 1.862 12.4 9 2.056 5.2 45 1.906 12.2 10 2.004 5.0 46 1.968 12.0 11 1.971 4.8 47 1.991 11.8 12 1.974 4.6 48 1.951 11.6 13 2.003 4.4 49 1.920 11.4 14 2.005 4.2 50 1.930 11.2 15 1.973 4.0 51 1.934 11.0 16 1.941 3.8 52 1.930 10.8 17 1.920 3.6 53 1.945 10.6 18 1.900 3.4 54 1.988 10.4 19 1.888 3.2 55 2.015 10.2 20 1.902 3.0 56 1.975 10.0 21 1.938 2.8 57 1.932 9.8 22 1.957 2.6 58 1.935 9.6 23 1.930 2.4 59 1.935 9.4 24 1.905 2.2 60 1.909 9.2 25 1.909 2.0 61 1.894 9.0 26 1.913 1.8 62 1.919 8.8 27 1.912 1.6 63 1.924 8.6 28 1.929 1.4 64 1.853 8.4 29 1.990 1.2 65 1.797 8.2 30 2.051 1.0 66 1.834 8.0 31 2.004 0.8 67 2.146 7.8 32 1.938 0.6 68 2.873 7.6 33 1.912 0.4 69 4.096 7.4 34 1.888 0.2 70 5.826 7.2 35 1.856 0.0 71 8.716 7.0 36 1.833

IN! lltlin Unit 1 Cycle 19 N 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 MWD/MTU Core Height Axial Unrodded Core Height Axial Unrodded (Ft.) Point Fxy (Ft.) Point Fxy 14.00 1 6.466 6.80 37 2.203 13.80 2 5.154 6.60 38 2.241 13.60 3 3.840 6.40 39 2.218 13.40 4 2.707 6.20 40 2.166 13.20 5 2.517 6.00 41 2.132 13.00 6 2.322 5.80 42 2.119 12.80 7 2.165 5.60 43 2.102 12.60 8 2.080 5.40 44 2.086 12.40 9 2.036 5.20 45 2.103 12.20 10 2.012 5.00 46 2.139 12.00 11 2.008 4.80 47 2.138 11.80 12 2.021 4.60 48 2.085 11.60 13 2.053 4.40 49 2.040 11.40 14 2.058 4.20 50 2.032 11.20 15 2.040 4.00 51 2.016 11.00 16 2.030 3.80 52 1.998 10.80 17 2.033 3.60 53 1.998 10.60 18 2.033 3.40 54 2.032 10.40 19 2.036 3.20 55 2.056 10.20 20 2.063 3.00 56 2.002 10.00 21 2.109 2.80 57 1.950 9.80 22 2.137 2.60 58 1.930 9.60 23 2.119 2.40 59 1.904 9.40 24 2.100 2.20 60 1.884 9.20 25 2.106 2.00 61 1.894 9.00 26 2.110 1.80 62 1.937 8.80 27 2.112 1.60 63 1.967 8.60 28 2.124 1.40 64 1.947 8.40 29 2.163 1.20 65 1.961 8.20 30 2.203 1.00 66 2.070 8.00 31 2.179 0.80 67 2.439 7.80 32 2.151 0.60 68 3.175 7.60 33 2.150 0.40 69 4.298 7.40 34 2.155 0.20 70 5.843 7.20 35 2.160 0.00 71 8.575 7.00 36 2.168