NOC-AE-13003058, Cycle 18 Core Operating Limits Report

Cycle 18 Core Operating Limits Report

ML13358A390
Person / Time
Site:	South Texas
Issue date:	12/05/2013
From:	Dunn R South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-13003058
Download: ML13358A390 (18)

Text

Nuclear Operating Company South Texas Pro/ectElectrIc GeneratlngStation P0. Box 289 Wadsworth, Texas 77483 -Avv---

December 5, 2013 NOC-AE-1 3003058 10CFR50.36 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 18 Core Operatingq Limits Report In accordance with Technical Specification 6.9.1.6.d, STP Nuclear Operating Company submits the attached Revision 1 to the Core Operating Limits Report for Unit 1 Cycle 18. The Report reflects the changes in Unrodded FxY for each core height from the updated Westinghouse ANC9 code system. The updated Fxy values are listed in Table 1 of the attachment.

There are no commitments in this letter.

If there are any questions regarding this report, please contact Ken Taplett at (361) 972-8416 or me at (361) 972-7743.

Roland F. Dunn Manager, Nuclear Fuel & Analysis web

Attachment:

Unit 1 Cycle 18 Core Operating Limits Report, Rev. 1 STI: 33793136

NOC-AE-13003058 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 1600 East Lamar Boulevard Arlington, TX 76011-4511 Balwant K. Singal U.S. Nuclear Regulatory Commission Balwant K. Singal Senior Project Manager John Ragan U.S. Nuclear Regulatory Commission Chris O'Hara One White Flint North (MS 8 B13) Jim von Suskil 11555 Rockville Pike NRG South Texas LP Rockville, MD 20852 Kevin Polio NRC Resident Inspector Richard Peia U.S. Nuclear Regulatory Commission City Public Service P.O. Box 289, Mail Code: MN116 Wadsworth, TX 77483 Peter Nemeth Crain Caton & James, P.C.

Jim Collins City of Austin C. Mele Electric Utility Department City of Austin 721 Barton Springs Road Austin, TX 78704 Richard A. Ratliff Robert Free Texas Department of State Health Services

MA M=

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

IE 3 Unit 1Cycle 18 Nuclear Operating Company Core Operating Limits Report Rev. 1 SAFMr- Page 2of 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.

.uE~l Unit 1 Cycle 18 ar Operating CoMpany Core Operating Limits Report Rev. 1 Page3 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

• l measured reactor vessel AT lead/lag time constant, -cl = 8 sec T2 measured reactor vessel AT lead/lag time constant, T2 = 3 sec T3 measured reactor vessel AT lag time constant, T3 = 2 see TU4 measured reactor vessel average temperature lead/lag time constant, 4=

tC 28 sec T5 measured reactor vessel average temperature lead/lag time constant, 'r5 = 4 sec Tr 6 measured reactor vessel average temperature lag time constant, 'r6 = 2 sec K1 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/psig T' Nominal full power Tavg, T'< 592.0 'F P' Nominal RCS pressure, P' = 2235 psig f1 (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%, fl(AI) = 0, where qt 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.

Over-power AT Setpoint Parameter Values Ti measured reactor vessel AT lead/lag time constant, TI = 8 sec

'72 measured reactor vessel AT lead/lag time constant, 'r2 = 3 sec T3 measured reactor vessel AT lag time constant, T3 = 2 sec T6 measured reactor vessel average temperature lag time constant, T6 = 2 sec T7 Time constant utilized in the rate-lag compensator for Tavg, -17 = 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 'F f2(AI) = 0 for all (Al)

ear perating ompanynit Cycle 18 lea t C Core Operating Limits Report Rev. 1 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, HRP 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).

T M NuclearOperatng Comnpany rUnit I Cycle 18 Core Operating Limits Report Rev. 1 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 FQRP = 2.55.

2.7.2 K(Z) is provided in Figure 7.

2.7.3 The Fxy limits for RATED THERMAL POWER (F 7) 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 PFy = 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-83 85. Therefore, these Fc 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 3.33.312, 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 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.

Unit 1 Cycle 18 Nuclear Operating Company Core Operating Limits Report Rev. 1 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 FX,(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 FAr = 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 (UFtH) to be applied to the FP' using the PDMS shall be the greater of:

UF&H = 1.04 OR UFrH = 1.0 + (UAH/I100)

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 the power distribution monitoring system.

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

UF, = 1.04 Applies to all fuel in the Unit I Cycle 18 Core.

Unit 1 Cycle 18 Nuclear Operat"iflgComlfy Core Operating Limits Report Rev. 1 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 Tav, -< 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 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 5Z529ZB01025 Rev. 4, Design Basis Document, Technical Specifications /LCO, Tech Spec Section 3.2.5.c 3.6 Document RSE-U 1, Rev. 1, "Unit 1 Cycle 18 Reload Safety Evaluation and Core Operating Limits Report". CR 12-23998-16 3.7 Letter from J. M. Ralston (Westinghouse) to D. F. Hoppes (STPNOC), "South Texas Project Electric Generating Station Unit 1 Confirmation of Requirements for the ANC9/BEACON 7 Mid-Cycle Implementation for South Texas Unit 1 (TGX) Cycle 18" NF-TG-13-73 (ST-UB-NOC-13003359) dated October 22, 2013 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.

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. 1 off mrPage 8 of16 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 (%)

N rtin PUnit 1 Cycle 18 Nuclear Operating Company Core Operating Limits Report Rev. 1

-Page 9 of 16 Figure 2 Required Shutdown Margin for Modes 3 & 4 7.0 6.0

- I. I 1  !- - .. I I - ..~....... I..... -I 1 L 1.L.... ....I -

- - iF- -- - F - I - i  !--- ! i - !I-- i - -I---- i 4- - F - - 4-!

Acceptable

-F--F-I -t - F- I 4- F -I- +-F-- 4-4--- -~ - + F -F---- ---- (2400,5.15) 5.0 4.0 3.0 10 2.0 L HH- +/-zLI- Hr+/-

i TTn2ccnt~h1~ L I

I! I m ,

I - T(600,1.30) 1.0 $.l I Il__

0.0

- - -- , 1 0 400 800 1200 1600 2000 2400 RCS Critical Boron Concentration (ppm)

(for ARI minus most reactive stuck rod)

Unit 1 Cycle 18 Nuclear Operating Company Core Operating Limits Report Rev. 1 Page 10 of 16 Figure 3 Required Shutdown Margin for Mode 5 7.0 6.0 5.0 4.0 I-

• - 3.0 cj_

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)

• &TMP Unit 1 Cycle 18 Nuclear Operating Company Core Operating Limits Report Rev. 1

-ENJ-- Page 11 of 16 Figure 4 MTC versus Power Level 7.0 6.0 Unacceptab'le 5.0

• ' 4.0 E

Acceptable 1 7;

0 3.0 20 T

02.0 -~ _ ~iij~I\

1{11771

-I-

" 1.0 0

-I-o.o

~ -_ _-~ \-

-1.0

-2.0 __ - - -~ __ __~-

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

I I M'l l ,Unit 1 Cycle 18 Nuclear Operating Company Core Operating Limits Report Rev. I off N Page 12 ofl16 Figure 5 Control Rod Insertion Limits* versus Power Level 260 I I--/- ( 23 ,259): 122 Step Overlap IFY (79 ,259): 122 Step Overlap

( 22,258): 121 Step Overlap. F- - -- 4-4 4 - I,'~- -- (78 258): 121 Step Overlap F 4-~

240 ' iz111 It. 1 I-I .. I I II I I I I II . I

.LJ 220 (020 B J I I I I I I I 200 180 II1 S1111 I i LI, nkC L 4

or It 11)

II c- 160 Ql)

I ~ ~~  !.. ~ I V..

.S 140

,,/__

-- z ' , 'i 'r

.A I II 120 100  ! !L/I, 80 60 controlb ar a t I o I 2 o 2 step 40 . . . . .I r3 I It i r

• i Control Bank A is already w~ithdra,.n to Full Out Position.

' : *I i *Fully writhdrawn shall be the condition where shutdown and 20

Il*control banks are at the position of either 258 or 259 steps I -,

(2 ) 0 -I 1 1 1 1 1 it hdra,,n .

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

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

-E N - Page 13 of 16 Figure 6 AFD Limits versus Power Level 120 I!1 1 -1TTT-'-T_

I I! I I I 1 I

"'III 110 100 " "I II I _

90

. . . [ I(

. . [1.

4+

if I Y]

-1 1 , 90 )Aclceptable II

( 11 , 9 0 )

I 1 11 1 1 80 Unacceptale I I nacceptable

-_ Acceptablee],

70 _

0 60 L

50 40 30 20 10 I~ L i ___IT 0

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

Unit 1 Cycle 18 lear Operating Company Core Operating Limits Report Rev. I Page 14 of 16 Figure 7 K(Z) - Normalize d FQ (Z) ve rs us Core Height 1.2 1.1 I:. . . . . . . . . . . .

1.0 i i i i  ; ii i i iI ii i i I i i i! 1 !..I I .... . . . .; I I  ! [;! I 0.9 i I . I ý . i .I . I. I I I. i iI !I II  !  ;  !  ; i i I I I I I I I I II I I I I 0.8 I ji ' Ii* IiI......i 1 iI111 I i I 111 11 11 f 0.7

.- 0.6 HilIIICore Elev. (it) FQ K(Z) 0.0 2.55 1.0 1 1 17.0 T 2.55 1.0

.t 0.5 1111 14.0 2.359 0.925 0.4 0.3 0.2 I

0.1 0.0 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 Core Height (ft)

Unit 1 Cycle 18 Nuear operatg Compay Core Operating Limits Report Rev. 1 Off A f t5Page 15 of 16 Table 1 (Part 1 of 2)

Unrodded Fy 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 11.503 6.80 37 1.894 13.80 2 8.767 6.60 38 1.930 13.60 3 6.030 6.40 39 1.923 13.40 4 3.573 6.20 40 1.888 13.20 5 2.768 6.00 41 1.869 13.00 6 2.407 5.80 42 1.872 12.80 7 2.237 5.60 43 1.880 12.60 8 2.156 5.40 44 1,888 12.40 9 2.079 5.20 45 1.915 12.20 10 2.029 5.00 46 1.956 12.00 11 2.002 4.80 47 1.971 11.80 12 2.001 4.60 48 1.937 11.60 13 2.026 4.40 49 1.899 11.40 14 2.032 4.20 50 1.909 11.20 15 2.008 4.00 51 1.913 11.00 16 1.984 3.80 52 1.920 10.80 17 1.967 3.60 53 1.940 10.60 18 1.950 3.40 54 1.983 10.40 19 1.937 3.20 55 2.019 10.20 20 1.944 3.00 56 1.969 10.00 21 1.972 2.80 57 1.933 9.80 22 1.993 2.60 58 1.927 9.60 23 1.966 2.40 59 1.927 9.40 24 1.946 2.20 60 1.932 9.20 25 1.950 2.00 61 1.968 9.00 26 1.955 1.80 62 2.044 8.80 27 1.959 1.60 63 2.067 8.60 28 1.969 1.40 64 1.997 8.40 29 1.980 1.20 65 2.001 8.20 30 2.012 1.00 66 2.076 8.00 31 1.993 0.80 67 2.442 7.80 32 1.960 0.60 68 3.213 7.60 33 1.940 0.40 69 4.694 7.40 34 1.915 0.20 70 6.578 7.20 35 1.889 0.00 71 7.318 7.00 36 1.872

• T M Unit I Cycle 18 N Oa Core Operating Limits Report Rev. 1 A N 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 3.618 6.80 37 2.111 13.80 2 3.456 6.60 38 2.160 13.60 3 3.294 6.40 39 2.153 13.40 4 3.111 6.20 40 2.126 13.20 5 2.798 6.00 41 2.111 13.00 6 2.407 5.80 42 2.106 12.80 7 2.163 5.60 43 2.098 12.60 8 2.067 5.40 44 2.086 12.40 9 2.030 5.20 45 2.099 12.20 10 2.034 5.00 46 2.133 12.00 11 2.020 4.80 47 2.133 11.80 12 2.020 4.60 48 2.078 11.60 13 2.053 4.40 49 2.031 11.40 14 2.072 4.20 50 2.025 11.20 15 2.053 4.00 51 2.013 11.00 16 2.035 3.80 52 1.997 10.80 17 2.029 3.60 53 1.992 10.60 18 2.032 3.40 54 2.015 10.40 19 2.031 3.20 55 2.048 10.20 20 2.055 3.00 56 1.981 10.00 21 2.097 2.80 57 1.905 9.80 22 2.118 2.60 58 1.874 9.60 23 2.100 2.40 59 1.847 9.40 24 2.087 2.20 60 1.822 9.20 25 2.090 2.00 61 1.815 9.00 26 2.087 1.80 62 1.838 8.80 27 2.081 1.60 63 1.857 8.60 28 2.079 1.40 64 1.826 8.40 29 2.097 1.20 65 1.835 8.20 30 2.132 1.00 66 1.905 8.00 31 2.101 0.80 67 2.119 7.80 32 2.075 0.60 68 2.442 7.60 33 2.066 0.40 69 2.662 7.40 34 2.065 0.20 70 3.184 7.20 35 2.067 0.00 71 5.344 7.00 36 2.074