ML061910067

Core Operating Limits Reports

ML061910067
Person / Time
Site:	South Texas
Issue date:	06/29/2006
From:	Leazar D South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-06002035
Download: ML061910067 (34)

Text

Nuclear Operating Company South T hosProed Electrnc Gencratin$Station PO. Box289 Wadsmorta exs 77483 June 29, 2006 NOC-AE-06002035 U. S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738 South Texas Project Units 1 and 2 Docket No. STN 50-498 and STN 50-499 Core Operating Limits Reports In accordance with Technical Specification 6.9.1.6.d, STP Nuclear Operating Company submits revised Core Operating Limits Reports for Unit 1 and for Unit 2. These revisions result from implementing the Westinghouse Best Estimate Analyzer for Core Operations -

Nuclear (BEACON) power distribution monitoring system as approved by the NRC for use at the South Texas Project (ML06070501).

There are no commitments in this letter.

If there are any questions regarding this request, please contact John Conly at (361) 972-7336 or me at (361) 972-7795.

ThU 4Z.. David A. Leazar Manager, Fuels & Analysis jtc Attachments: Unit 1 Cycle 13 Core Operating Limits Report, Revision 1 Unit 2 Cycle 12 Core Operating Limits Report, Revision 1 Aool STI: 32022440

NOC-AE-06002035 Page 2 of 2 cc:

(paper copy)

(electronic copy)

Regional Administrator, Region IV U. S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, Texas 76011-8064 Richard A. Ratliff Bureau of Radiation Control Texas Department of State Health Services 1100 West 49th Street Austin, TX 78756-3189 Senior Resident Inspector U. S. Nuclear Regulatory Commission P. 0. Box 289, Mail Code: MN16 Wadsworth, TX 77483 C. M. Canady City of Austin Electric Utility Department 721 Barton Springs Road Austin, TX 78704 A. H. Gutterman, Esquire Morgan, Lewis & Bockius LLP Mohan C. Thadani U. S. Nuclear Regulatory Commission Steve Winn Christine Jacobs Eddy Daniels NRG South Texas LP J. J. Nesrsta R. K. Temple E. Alarcon City Public Service Jon C. Wood Cox Smith Matthews C. Kirksey City of Austin

• A

,Nuclear:Operating :Company.:.!.

SOUTH TEXAS PROJECT Unit 1 Cycle 13 CORE OPERATING LIMITS REPORT Revision 1 Core Operating Limits Report Page I of 16

TUnit I Cycle 13 N

O Core Operating Limits Report Rev. I Pane2 of 16 LO.

CORE*OPERATING LIMITS REPORT This Core Operating Limits Report for STPEGS Unit 1 Cycle 13 has been prepared in accordance with the requirements of Technical Specification 6.9.1.6." The core operatinglimits ha('e 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).

3),

4) 5)

6) 7)8) 9)

10) 2.1 2.2 3/4.1.1.1 3/4.1.1.3 3/4.1.3.5 3/4.1.3.6 3/4.2.1 3/4.2.2 3/4.2.3 3/4.2.5 SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS SHUTDOWN MARGIN MODERATOR TEMPERATURE COEFFICIENT LIMITS SHUTDOWN ROD INSERTION LIMITS CONTROL ROD INSERTION LIMITS AFD LIMITS HEAT FLUX HOT CHANNEL FACTOR NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR 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 M Unit 1 Cycle 13 N

n any Core Operating Limits Report Rev. 1 Page 3 of 16 2.2.2 The Over-temperature AT and Over-powerAT setpoint parameter values are listed below:

Over-temperature AT Setpoint Parameter Values'.

T" measured reactor Vessel AT.lead/lag time constant, ", = 8 see c2 measured reactor vessel AT-lead/lag time constant, "2 = 3 see "c3 measured reactor vessel AT lag time constantT 3 = 2 see c4 measured reactor vessel average temperature lead/lag time constant, T4 = 28 see "c5 measured reactor vessel average temperature lead/lag time constant, 'Z = 4 see c6 measured reactor vessel average temperature lag time constant, E6 = 2 see K1 Overtemperature AT reactor trip setpoint, K1 =114 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 Tayg, T'_5 592.0 OF 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 star~tup tests such that:

(1)

For q, - qb between -70% and +8%, f1(At) = 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 TPOWER;

• (2)

For each percent that the magnitude of q, - qb exceeds -70%, the AT Trip Setp6int 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 Tr measured reactor vessel AT lead/lag time constant, cl = 8 see "T2 measured reactor vessel AT lead/lag time constant, r2 = 3 see c3 measured reactor vessel AT lag timeconstant, 73 = 2 see 76 measured reactor vessel average temperature lag time constant, T6 = 2 see T7 Time constant utilized in the rate-lag compensator for Tavg, ¶7 = 10 see K4 Overpower AT reactor trip setpoint, K4 = 1.08 Ks Overpower AT reactor trip setpoint Tavg rate/lag coefficient, KS = 0.02/7F for increasing average temperature, and K5 = 0 for decreasing average temperature 1K6 Overpower AT reactor trip setpoint Ta,,g heatup coefficient K6 = 0.002/7F for T>T",and K6 = 0forT_5 T" T"

Indicated full power T,,, T"_5 592.0 OF ft(A1) = 0 for all (AI)

T l

Unit 1 Cycle 13 NuclearOprating Company Core Operating Limits Report Rev. I

'N Page4 of 16

...2.3 SHUTDOWN.MARGIN (Specification 3.1.i):

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 limitsin 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 pcn/IF.

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 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 benchmuark data contained in the surveillance procedure are met, then an MTC measuriement 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 at least 250 steps withdrawn but not exceeding 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 iM Unit 1 Cycle 13 Nuclear Operating Company Core Operating Limits Report Rev..1 O N A

Page 5 of 16 2.6 AX" IL FLUX DIFFERENCE (Specification3.2.1):

2.61 AFD limits as required by Technical Specification 3.2.1 are determined by 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 Fxy limits for RATED THERMAL POWER (FR} ) 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 Fy(Z) using the PDMS shall be calculated by:

UFQ = (1.0 + (UQ/lOO))*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.

WT M Unit 1 Cycle 13 Nuclear Operating Company Core Operating Limits Report Rev. I APage 6 of 16 2.7.4.2' Ifthemoveable detector-system is used, the core power distribution measurement uncertainty.(UFQ) to be applied to the FQ(Z) and Fy(Z) shall be calculated by:

UFQ =UQU*UE Where:

UQU =Base FQ measurement uncertainty of 1.05.

UE=

Engineeiing uncertainty factor. of 1.03.

2.8 ENTHALPY RISE HOT CHANNEL FACTOR (Specification 3.2.3):

I 2.8.1 FART

= 1.621 2.8.2 PFH = 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) isoperable, as defined in the Technical Requirements Manual, the core power distribution measurement uncertainty (UFAH) to be applied to the Fa using the PDMS shall be calculated by:

UFAH = 1.0 + (UAH/100)

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 (UFWa) shall be:

Ur-H = 1.04 Applies to all fuel in the Unit 1 Cycle 13 Core.

T M Unit 1 Cycle 13 Nuclear Operating Conipany Core Operating Limits Report Rev. I

-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 Tayg, _5 595 'F 2 2.9.1.2

Pressurizer Pressure,. > 2200.psig 3,

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

3.0 REFERENCES

3.1 Letter from D. E..Robinson (Westinghouse) to D. F. Hoppes (STPNOC), "Unit 1 Cycle 13 Final Reload Evaluation (RE) Revision 1," ST-UB-NOC-05002532, Rev. 1, March 29 2005.

3.2 NUREG-1346, Technical Specifications, South Texas Project Unit Nos. 1 and 2.

3.3.

STPNOC Calculation ZC-7035, Rev. 2, "Loop Unc&rtainty Calculation for RCS Tavg hInstmentation," Section 10.1, effective July22, 2003.

3.4 STPNOC Calculation ZC-7032, Rev. 4, "Loop Uncertainty Calculation for Narrow Range Pressurizer Pressure Monitoring Instrumentation," Section 2.3, Page 9, effective July 22, 2003.

3.5 Condition Report Engineering Evaluation 03-6461-9, Revision-0, "Reload Safety Evaluation and

• Core Operating.Limits Report for South Texas Unit 1 Cycle 13 Modes 1, 2, 3, 4, and 5."

3.6 WCAP-12472-P-A, BEACON Core Monitoring and Operations Support System, August 1994.'

I 1 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 IF 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.

Nuclear Operating Company ON AN-Unit I Cvcle 13 Core Operating Linmits Report Rev. 1 Page 8 of 16 Figure 1.

Reactor Core Safety Limits - Four Loops in Operation 680 660 640 620 0

U 600 1171171T17.[1iT1771 I I KITF7fFTfI7~7.I I I 1 t I t I

Unacceptable

~2_450 P 7SIA]_

• < ~12250 Ps-S IA

-.. e

[98

.637.981--j_

(102.622.62)

-7**,*

I HID 580 560 540 1 10.12,587135)

Acceptable:

1(13 'S-I 0

20 40 60 80 Rated Thermal Power (%)

100 120 140

Unit 1 Cycle 13 Nuclear Operating Company Core Operating Limits Report Rev. 1 W

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

~z.z1 11 4 4 i~i~

4

-f Acceptable i-i s--I-

-ii

-i -

i-i t

k -

2400.5.15)1

-r -i-i-r -i-

~1~

t r -

r -

1~

r -

r -

7 1~

71 r

i 5.

t%

"U I I 4.0 3.0 rd C4

-II.

i/

I I

SUnacceptabl.-

S I (60,130) 2..0 F-1.0 0.0 0

400 800 1200 1600 RCS Critical Boron Concentration (ppm)

(for ARI minus most reactive stuck rod) 2000 2400

~VM*.

Unit 1 Cycle.13

,ctsar O~erating Company" Core Operating Limits Report Rev. I

'":rPage 10 of 16

.Figure 3 Required Shutdown Margin for Mode 5 7.0 6.0 Acceptable I I.

I 5.0 S4.0 0

• _ 3.0 2=

2).0 I

.I I I: I I

i i

2 0 J

)

.4.50)1

.F I

1 4-4-4-4-4 4-4-4-4 -

1 (0,1.50) 1 0

Unacceptablej TI hTI I I

I I

-- 0 50*.,.3o )

1.0 0.0 II I

II I-0 400 800 1200 1600 RCS Critical Boron Concentration (ppm)

(for ARI minus most reactive stuck rod) 2000 2400

Unit 1 Cycle 13 Nuclear Operating Company NlMa pran C Core Operating Limits Report Re%,. I Page 11 of 16 Figure 4 MTC versus Power Level 7.0 6.0

.:*I

.-I -

I 5.0 4.0 0

• . 3.0 C.

1.0 1.

bo 0.0

-1.0 Unacceptable Operation S-cceptable Operation1]

I j

-2.0

-3.0 0

10 20 30 40 50 60 Rated Thermal Power (%)

70 80 90 100

jT M.

Unit 1 Cycle 13 Nuclear Operating Company Core Operating Limits Report Rev. I APage 12 of 16 XFigure 5 Control Rod Insertion Limits* versus Power Level 260 240 220 200 180 till I Y (23,259):

(23,258 ):

(22,256):

21,254):

(20,252):

(19,250):

122 Step Overlap 121 Step Overlap (79 (79 119 117 115 113 Step.Overlap I-I I 1 7 (78 Step Overlap I

l

( 7I Step Overlap _

1 1 1

i t (7

9.,259): 122 Step Overlap 9,258): 121 Step Overlap

,256): 119 Step Overlap 254 ):. 117 Step Overlap 6,252.): 115 Step Overlap 5,250 ): 113 Step Overlap 160 140 120 100 I

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-T I-T Control Bank A is already withdrawn to Full Out Position.

SFully withdrawn region shall be the condition where shutdown and control banks are at a position within the interval of 250 F~F

-and less than or equal to 259 steps withdrawn, inclusive.

80 60 40 20 0

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

f l fl Unit 1 Cycle 13 Nuclear Operatng Company Core Operating Limits Report Rev.

SO N A Page 13 of 16 Thi*rnn 6

-a---

120 110.

100 90 80 Ca 0*

70 60 50 40 30 20 10 0

AMl) Limits versu PowerLevel L.

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

Unit 1 Cycle 13 Nuclear Operating Company Core Operating Limits Report Rev. 1 Offr A Pane 14 of16 Figure 7 K(Z) - Normalized FQ (Z) versus Core Height

-1.2 1.1 1.0 0.9 0.8

• 0.7

• 0.6

.t* 0.5 z

0.4 03.3 0.2 0.1 0.0 1 1 1 1 1 1 l i l t l i l t I I I l i l t l i l t 1 1 1 1 1 1 1 1 1 f i l l I f 1 1 I I I It I I I 1 1.1 7

I I I I if ALLL III I

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_U I.A lilt I I It-I - I...

1111 1 1 TFTE Tm-I -L I lilt till lilt I-1111 lilt 1=111 4 4-H+ 444' I I I l i l t l i l t t i l l 1 1 1 t i l l 1 1 1 l i l t I I 1 1 1 1 1 1 1 1 1 1 1 11 1 1 1 1 1 1 I I I I I I f i l l I I H

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=

FTEE I l i l t 1 1 1 1 1 1 1 1 1 FT-F 11 1 1 i-1

.1 Core Elev. (ft) 0.0 7.0 14.0 FQ.*

2.55 2.55 2.359 K(Z) 1.0 1.0 0.925 1:1 I

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1-11 1. LLI 11 1 1-1111 1-1.1 1 -I-LLI

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0 1

2 3

5 6

7 8

Core Height (ft) 9 10 11 12 13 14

Unit 1 Cycle 13 Nuclear 0erating Company Core Operating Limits Report Rexv. I SAPage 15 of 16 Table 1 (Part 1 of 2)

Unrodded Fy for Each Core'Height for Cycle Burnups Less Than 10500 MWD/MTU Core Height (Ft.)

Axial Point Unrodded Fxy Core Height Axial:

(Ft.)

Point Unrodded Fxy*

[

14.00 13.80 13.60 13.40 13.20 13.00 12.80 12.60 12.40 12.20 12.00 11.80 11.60 11.40 11.20 11.00 10.80 10.60 10.40 10.20 10.00 9.80 9.60 9.40 9.20 9.00 8.80 8.60 8.40 8.20 8.00 7.80 7.60 7.40 7.20 7.00 1

2 3

4 5

6 7

8 9

10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 4.633 3.954 3.275 2.596 2.276 2.068 2.070 2.046 2.023 1.995 1.979 1.971 1.971 1.975 1.979 1.984 1.984 1.984 1.982 1.985 1.987 1.989 1.997 2.003 2.011 2.017 2.023 2.027 2.032 2.037 2.033 2.016 1.983 1.945 1.911 1.889 6.80 6.60 6.40.

6.20 6.00 5.80 5.60 5.40 5.20 5.00 4.80 4.60 4.40 4.20 4.00 3.80 3.60 3.40 3.20 3.00 2.80 2.60 2.40 2.20 2.00 1.80 1.60 1.40 1.20 1.00 0.80 0.60 0.40 0.20 0.00 37 38 39 40 41 42 43 44 45.

46

• .47 48 49 50 52 53 54..

55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71

• 1.872 1.861 1.853 1.845 1.841 1.840 1.839 1.840 1.844 1.850

.1.860 1.870 1.878 1.885 1.890 1.893 1.893 1.895 1.899 1.904 1.903 1.911 1.921 1.932 1.929 1.920 1.905 1.907 1.909 1.920 1.994 2.124 2.281 2.439 2.596

Ar/F_.

Unit 1 Cycle 13 Serating Company Core Operating Limits Report Rev. 1 Page 16 of] 6 Table l.(Part2of2)

Unrodded Fxy for Each Core Height for Cycle Burnups Greater Than or Equal to 10500 MWrD/MTU Core Height (Ft.)

Axial P6int.

-Unrodded Fxy' Core Height Axial Point Unrodded

. :Fxy 14.00 13.80 13.60 13.40 13.20 13.00 12.80 12.60 12.40 12.20 12.00 11.80 11.60 11.40 11.20 11.00 10.80 10.60 10.40 10.20 10.00 9.80 9.60 9.40 9.20 9.00 8.80 8.60 8.40 8.20 8.00 7.80 7.60 7.40 7.20 7.00

• 1..

2

.3 4

5 6

78 9

• 10 11 12 13 14 15

.16 17 18 19 20 21 22 23.

24 25 26 27T 28 29 30 31 32 33 34 35 36 4.778

.4.129 3.480 2.831 2.450

.2.153 2.149 2.121

• 2.096 2.'055 2.03.2 2.026 2.019 2.022 2.027 2.031Y 2.033.

2.035 2.036 2.041 2.048 2.056 2.065.

2.072 2.078 2.084 2.089 2.095 2.100 2.106 2.112 2.119 2.127 2.135 2.142 2.145 I

6.80 6.60 6.40 6.20 6.00 5.80 5.60 5.40 5.20

• 5.00 4.80 4.60 4.40 4.20 4.00:

3.80 3.60 3.40 3.20

.3.00 "2.80 2.60 2.40 2.20 2.00 1.80 1.60 1.40 1.20 1.00 0.80 0.60 0.40 0.20 0.00

.37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52

.*53' 54.

55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 2.144 2.139 2.127 2.115 2.101 2.089 2.077 2.067 2.057 2.047 2.038 2.028 2.017 2.006 1.994 1.982 1.969 1.957 1.945 1.931 1.916 1.891 1.858 1.841 1.839 1.8 35 1.836 1.853 1.853 1.877 2.127 2.605 3.195 3.786 4.377

Nuclear Operating Company SOUTH TEXAS PROJECT Unit 2 Cycle 12 CORE OPERATING LIMITS REPORT Revision 1 Core Operating Limits Report Page I of 16

r il AN Unit 2 Cycle 12 Nuclear operating Company Core Operating Limits Report Re'v. I Page 2 of 16 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report for.STPEGS Unit 2 Cycle 12 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 (Tvg) 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.

0

,,T Unit 2 Cycle 12 Nuclear Operating Company Core Operating Limits Report Rev. 1 Pa 3 of16 2.2.2 The Over-temperature AT anid Over-power AT setpoint parameter values arelisted below:

Over-temperaturie. AT Setpoint Parameter Values Ti measured reactor vessel AT lead/lag time constant, ri = 8 see E2 measured reactor vessel AT lead/lag time constant, T2 = 3 see c3 measured reactor vessel AT lag time constant, c3 = 2 see T4 measured reactor vessel average temperature lead/lag time constant, c4 = 28 see T5 measured reactor vessel average temperature lead/lag time constant, 'r5 = 4 see T6 measured reactor vessel average temperature lag time constant, T6 2 see K,

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, K3 = 0.00143/psig T'

Nominal full power Tavg, T':5 592.0 OF 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 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.

Over-power AT Setpoint Parameter Values T"

measured reactor vessel AT lead/lag time constant, cl = 8 see

• 2 measured reactor vessel AT lead/lag time constant, c2 =3 see

'c3 measured reactor vessel AT lag time constant, "3 = 2 see

'T6 measured reactor vessel average temperature lag time constant, -,6 =2 see T7 Time constant utilized in the rate-lag compensator for Tavg, T7 = 10 see K4 Overpower AT-reactor trip setpoint, K4 = 1.08 K5 Overpower AT reactor trip setpoint Tayg rate/lag coefficient, Ks = 0.02/°F for increasing average temperature, and Ks = 0 for decreasing average temperature K6 Overpower AT reactor trip setpoint Tavgheatup coefficient K6 = 0.002/°F for T>T",and K6 = 0forT< T" T"

Indicated full power Tag, T"_5 592.0 OF f2(AI) = 0 for all (AI)

dP WA--

Unit 2 Cycle 12 Nulear Operating Company Core Operating Limits Report Rev. I Page 4 of 16

.2.3 SHUTDOWN MARGIN (Specification 3.1.1.1):

The SHUTDOWN NMARGIN 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 T.S. 6.9.1.6.b.10:

Revised Predicted MTC = Predicted MTC + AFD Correction - 3 pcm/rF 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 either 255 steps withdrawn 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).

Unit 2 C-ycle 12 NuerOertnCopn Core Operating Limits Report Rev. I 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.CAOC Operations with an AFD target band of +5, -10%.

.2.6.2*.The AFD shall be maintained within the ACCEPTABLE OPERATIONportion of Figure 6, as required by Technical Specifications.

2.7 HEAT FLUX HOT CHANNEL FACTOR (Specification 3.2.2):

2.7.1 FR'

= 2.55.

2.7.2' K(Z) is provided in Figure 7.

2.7.3 The Fy limits for RATED THERMAL POWER (FRyT) 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 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, the core power distribution measurement uncertainty (UFQ) to be applied to the FQ(Z) and F,-,(Z) using the PDMS shall be calculated by:

2.7.4.2 UFQ = (1.0 + (TQ/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.

Unit 2 Cycle 12 uclear Operating Company Core Operating Limits Report Rev,. 1 J

• Page 6 of 16 2.7.4.3 If the moveable detectorsystem is used, the core power distribution measurement uncertainty (UFQ) to be applied to the FQ(Z) and Fy(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 FAT = 1.621 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, the core power distribution measurement uncertainty (UFAH) to be applied to the FM using the PDMS shall be calculated by:

UFAH = 1.0 + (UAH/100)

Where:

U

= Uncertainty for power peaking factor as defined in Equation 5-19 of Reference 3.6 This uncertdinty 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 (UFH) shall be:"

UF*H= 1.04 I

I Applies to all fuel in the Unit 2 Cycle 12 Core.

T M Unit 2 Cycle 12 Near Operating Company Core Operating Limits Report Rex'. I

' ar-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 Ta,,g, _<595 7F 2 2.9.1.2 Pressurizer Pressure, > 2200 psig,

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

3.0 REFERENCES

3.1 Letter from D.E. Robinson (Westinghouse) to D. F. Hoppes (STPNOC), "South Texas Project Nuclear Operating Company South Texas Project Electric Generating Station Unit 2 Cycle 12 Final Reload Evaluation (RE) Revision 2," ST-UB-NOC-05002587, Rev. 2, October 21, 2005.

3.2 NUREG-1 346, 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 Calculatiofi for Narrow Range Pressurizer Pressure Monitoring Instrumentation," Section 2.3, Page 9.

3.5 Condition Report Engineering Evaluation 04-5927-9, Revision 0, "Reload Safety Evaluation and Core Operating Limits Report for South Texas Unit 2 Cycle 12."

3.6 WCAP-12472-P-A, BEACON Core Monitoring and Operations Support System, August 1994.

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.

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.

dL,_J

!Unit 2 Cvcle 12 Nuctear Operating Company Core Operating Limits Report Rev. 1 rPage 8 of 16 Figure 1 680 Reactor Core Safety Limits - Four Loops in Operation It -LL i~-1~-1 1..

~64~j I

I IUnacc eptable 660 640-

__ 620 0

600 (2.652.10)62 It8.3

).1i) 580 560 540 112 II 173 I Acceptable (130..563.--

-3 0

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

M n,*i~ili Unit 2 Cycle 12 n,

Core Operating Linfits Report Page 9 of 16 Figure 2 Required Shutdown Margin for Modes 3 & 4 7.0 6.0 5.0 S4.0 S3.0 2.0 2.0 F--

111

-1 Acceptable i

-A-

+ -

~

I I -

t -

I -

1-11 1-1-I-t -

I -

I -

t

( 40 240,.1

.i 7

Unacceptablei

-I--I.

I I

I I--

1.0 0.0 I(0 3

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

(for ARI minus most reactive stuck rod) 2000 2400

TM

~

Unit 2 Cycle 12 Nuclear Operating Company Core Operating Limits Report Rex'. I Page 10 of 16 Figure 3 Required Shutdown Margin for Mode 5

-7.0 6.0 5.0 S4.0 13.0 rn~

2-.

2.0

-' Acceptable-

-".-T

-iv AF A-H,+-I--ff-A4.50) l 174tILl I

I 1 II 0

I( =Dt3)

-I-I

-I I

I Unacceptab1~j

-1I III m

I I

I I

I 650.1.30 1.0 0.0 mzzz~z~z~zx~

-~-.

T 4 1~

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

(for ARI minus most reactive stuck rod) 2000 2400

Unit 2Ccle, 12 Nuar Operatng Company Core Operating Limits Report Rev. 1 Page 11 of 16 Figure 4 MTC versus Power Level 7.0T

..I I I I I

.1 '1I 6.0 5.0 4.0 3.0 4!

U 2.0 1.0 0.

~0.0

-1.0

-2.0

-3.0

... I-----11I11-1 I

I3/4--

-tacceptable Operation]

Accptbl i

0 10 20 30 40 50 60 Rated Thermal Power (%)

70 80 90 100

f jlfl Unit2 Cycle 12 Nuclear Operating Company Core Operating Limits Report Rev. I W

"A*W M--

Page 12 of 16 Figure 5 Control Rod Insertion Limits* versus Power Level 260 240 220 200 180 j 160 140

" S 120 100 80 60 40 20 0

I.-

I I-j(2II I5_):

122 Step Overlap I(

1 179,259

'122 Step Overlap 1 (21,255): 118 Step Ovela I I 8St1 Ovelap týLal g-B!tN I

I I

I I

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' I I I I L OI 1!-

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I:

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30 1

5 60 70 Rt Tel Power (%)

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I

uTMaUnit 2 Cycle 12-Nia Or Core Operating Limits Report Rev. I Page 13 of 16 Figure 6 AFD Limits versus Power Level 120 110

• 1 100*

90 i l 80 Unacceptable Unacceptable Operation Acceptabl-I Operation S 7 0 -

A-

-c e pt,*

a"' l

°~

IJ I I 1 11!

m~

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/

I I I I Operation I I

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~

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iiiii 1~~i

-50

-40

-30

-20

-10 0

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

Unit 2 Cy-ce 12 Nu Or

.n Core Operating Limits Report Rev. I Page 14 of 16 1.2 Li.

1.0 0.9 0.8 0.7

• 3= 0.6.

Z 0.4 C,e 0.3 0.2 0.1 0.0 Figure.7-K(Z) - Normalized FQ(Z) versus Core Height II III 0 ~~~

JI 21 3

15 6

1 8

9 10 11 12 13 1

Core eitl I I I

i 1

1 1

1 1

I I

i i

l 1

1 7T I

T II I I i

l 1

1 1

1 1

1 J

l 1

1 11 1

4 Core Elev. (fI)

FQ K(Z)It t" I'

_0.0 2.55 1.0 I"- IIH I4 7.0 2.55 1.0

-14.0 2.359 0.925 1 1 1 6

7 8111 12 13 114F

-Ir IIeigh I(It)

Ar

~ tn g l

l om p

.Unit 2 Cycle 12 NuclarOperating Company Core Operating Limits Report Rexv. 1

~E Page 15 ofl16 Table 1 (Part.1 of 2)

Unrodded F*, for Each Core.Height for Cycle Burnups Less Than 9000 MWrD/MTU

.Core Height (Ft.)

Axial Point Unrodded Fxy I.Core Height Axial (Ft.)

-Point Unrodded Fxy 14.00 13.80 13.60 13.40 13.20 13.00 12.80 12.60 12.40 12.20 12.00 11.80 11.60 11.40 11.20 11.00 10.80 10.60 10.40 10.20 10.00 9.80 9.60 9.40 9.20 9.00 8.80 8.60 8.40 8.20 8.00 7.80 7.60 7.40 7.20 7.00 1

2 3

4 5

6 7

8 9

10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 34 35 36 5.271 4.497 3.723 2.949 2.563 2.303 2.269 2.209 2.156 2.104 2.073 2.053

.2.043 2.037

.2.033 2.029 1.023 2.016 2.010

• 2.010 2.010 2.010 2.014 2.017 2.021 2.026 2.034 2.043 2.054 2.065 2.073 2.069 2.054 2.038 2.010 1.976 6.80 6.60 6.40 6.20 6.00 5.80 5.60 5.40

• 5.20

• 5.00 4.80 4.60 4.40 4.20.

4.00

• 3.80 3.60 3.40

.3'20 3.00

.2.80 2.60 2.40

.2.20 2.00 1.80 1.60 1.40 1.20 1.00 0.80 0.60 0.40 0.20 0.00 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 54' 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 1.955 1.948 1.940 1.933 1.928 1.927 1.927 1.928 1.933 1.938 1.946 1.955 1.960 1.965 1.968 1.965 1.958 1.954 1.952 1.952 1.945 1.944 1.944 1.945 1.948 1.948 1.945 1.957 1.980 2.03 1 2.229 2.756 3.447 4.138 4.829

T MeaUnit 2 Cycle 12 ear O

.rtn CCore Operating Limits Report Rev. I Page 16 of 16 Ta'ble - (Part 2.of2)

Unrodded Fxy for Each Core Height for Cycle Burniups Greater Than or Equal to 9000 MWD/MTU Core Height Axial Unrodded Core Height Axial Unrodded (Ft.)

Point Fxy (Ft.)

Point:

Fxy 14.00 1

4.128 6.80 37 2.142 13.80 2

3.681 6.60 38

..2.141

.13.60

3.

3.234 6.40 39 2.129 13.40 4

2.788 6.20 40 2.116 13.20 5

2.497 6.00 41 2.102 13.00 6

2.254 5.80 42 2.092 12.80 7

2.190 5.60 43 2.081 12.60 8

2.133 5.40 44 2.071 12.40 9

2.084 5.20 45 2.060 12.20 10 2.043 5.00 46 2.051 12.00 11 2.014 4.80 47 2.042 11.80 12 2.013 4.60 48 2.035 11.60 13 2.017 4.40 49 2.025 11.40 14 2.026 4.20.

50 2.015 11.20 15 2.031 4.00 51 2.003 11.00 16 2.036 3.80

..52 1.992 10.80 17 2.042 3.60 53 1.982 10.60 18 2.045 3.40 54 1.972 10.40 19 2.046.

3.20 55 1.960

.10.20 20 2.048 3.00 56 1.944

• 10.00

. 21 2.049

.2.80 57 1.925 9.80 22 2.051 2.60 58 1.899 9.60 23

.2.054 2.40 "

59

..1.873 9.40 24 2.058 2.20 60 1.846 9.20 25 2.062 2.00 61 1.837 9.00 26 2.066 1.80 62

• 1.833 8.80 27 2.069 1.60 63 1.832 8.60 28 2.073 1.40 64 1.832 8.40 29 2.078 1.20 65 1.862 8.20 30 2.084 1.00 66 1.932 8.00 31 2.091 0.80 67 2.144 7.80 32 2.099 0.60 68 2.520 7.60 33 2.107 0.40 69 2.976 7.40 34 2.116 0.20 70 3.433 7.20 35 2.127 0.00 71 3.889 7.00 36 2.137