NOC-AE-15003251, Cycle 18 Core Operating Limits Report

Cycle 18 Core Operating Limits Report

ML15127A165
Person / Time
Site:	South Texas
Issue date:	04/26/2015
From:	Dunn R F South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-15003251
Download: ML15127A165 (18)

Text

JLMr/ll Nuclear Operating Company SouTexas Pro/e ectfdctric Generadn$Sttlon PWI .POx Wdsworth, 7Txas 74481 r April 28, 2015 NOC-AE-1 5003251 10 CFR 50.36 U.S. Nuclear Regulatory Commission Attention:

Document Control Desk Washington, DC 20555-0001 South Texas Project Unit 2 Docket No. STN 50-499 Unit 2 Cycle 18 Core Operating 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 2 Cycle 18. The report covers the core design changes made during the 2RE17 refueling outage.There are no commitments in this letter.If there 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 2 Cycle 18 Core Operating Limits Report, Revision 0 kooL STI: 34115800 NOC-AE-1 5003251 Page 2 of 2 cc: (paper copy)Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 1600 East Lamar Boulevard Arlington, TX 76011-4511 Lisa M. Regner Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North (08H04)11555 Rockville Pike Rockville, MD 20852 NRC Resident Inspector U. S. Nuclear Regulatory Commission P. O. Box 289, Mail Code: MN116 Wadsworth, TX 77483 (electronic copy)Morgan, Lewis & Bockius LLP Steve Frantz, Esquire U.S. Nuclear Regulatory Commission Lisa M. Regner NRG South Texas LP John Ragan Chris O'Hara Jim von Suskil CPS Energy Kevin Polio Cris Eugster L. D. Blaylock 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 2 Cycle 18 CORE OPERATING LIMITS REPORT Revision 0 Core Operating Limits Report Pa-ge I of 16 A MrerUnit 2 Cycle 18 Nuclear Operating Company Core Operating Linits Report Rev. 0 Page 2 of 16 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report for STPEGS Unit 2 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.

Unit 2 Cycle 18 Nuclear Operating Company Core Operating Limits Report Rev. 0 ff -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 T2 measured reactor vessel AT lead/lag time constant, Ti = 8 sec T2 measured reactor vessel AT lead/lag time constant, T2 = 3 sec T3 measured reactor vessel AT lag time constant, 2T3 = 2 sec T4 measured reactor vessel average temperature lead/lag time constant, T4 = 28 sec T5 measured reactor vessel average temperature lead/lag time constant, 2T5 = 4 sec T6 measured reactor vessel average temperature lag time constant, T6 = 2 sec Ki Overtemperature AT reactor trip setpoint, KI = 1. 14 K2 Overtemperature AT reactor trip setpoint Tavg coefficient, K 2 = 0.028/7F K3 Overtemperature AT reactor trip setpoint pressure coefficient, K3 = 0.00143/psi T' Nominal full power Tavg, T'< 592.0 'F P' Nominal RCS pressure, P' = 2235 psig fi(AI) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that: (1) For qt -qb between -70% and +8%, f 1 (AI) = 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 -qh 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 T2 measured reactor vessel AT lead/lag time constant, T2 = 3 sec T3 measured reactor vessel AT lag time constant, T3 = 2 sec T6 measured reactor vessel average temperature lag time constant, T6 = 2 sec T7 Time constant utilized in the rate-lag compensator for Tavg, T7 = 10 sec K4 Overpower AT reactor trip setpoint, K4 = 1.08 K5 Overpower AT reactor trip setpoint Tavg rate/lag coefficient, K5 = 0.02/7F 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 = OforT< T" T" Indicated full power Tavg, T"< 592.0 'F f 2 (AI) = 0 for all (Al)

Unit 2 Cycle 18 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 4 of 16 2.3 SHUTDOWN MARGIN (Specification 3.1.1.1): The SHUTDOWN MARGIN shall be: 2.3.1 Greater than 1.3% Ap for MODES 1 and 2**See Special Test Exception 3.10.1 2.3.2 Greater than the limits in Figure 2 for MODES 3 and 4.2.3.3 Greater than the limits in Figure 3 for MODE 5.2.4 MODERATOR TEMPERATURE COEFFICIENT (Specification 3.1.1.3): 2.4.1 The BOL, ARO, MTC shall be less positive than the limits shown in Figure 4.2.4.2 The EOL, ARO, HFP, MTC shall be less negative than -62.6 pcrn/°F.2.4.3 The 300 ppm, ARO, HFP, MTC shall be less negative than -53.6 pcmr/°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 algorithmn from 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).

W llMl M Unit 2 Cycle 18 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 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 Teclhical Specifications.

2.7 HEAT FLUX HOT CHANNEL FACTOR (Specification 3.2.2): 2.7.1 F~VP = 2.55.2.7.2 K(Z) is provided in Figure 7.2.7.3 The Fxy limits for RATED THERMAL POWER (FRyTP) witlhin 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 F, 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 F.y(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 fiom the document referenced by Technical Specification 6.9.1.6.b.

1I UE = Engineering uncertainty factor of 1.03.This uncertainty is and applied automatically by the Power Distribution Monitoring System (PDMS).

A lPlrAN Unit 2 Cycle 18 Nuclear Operating Company Core Operating Linits 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 F,,y(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 FP= 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 Fý, using the PDMS shall be the greater of: UFAH = 1.04 OR UFAH = 1.0 + (UA/100)Where: UAH = Uncertainty for power peaking factor as defined in Equation 5-19 from the document referenced 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 VnIII M Unit 2 Cycle 18 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 7 of 16 2.9 DNB PARAMETERS (Specification 3.2.5): 2.9.1 The following DNB-related parameters shall be maintained within the following limits: 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

firom J. M. Ralston (Westinghouse) to D. F. Hoppes (STPNOC), "South Texas Project Electric Generating Station Unit 2 Cycle 18 Final Reload Evaluation" NF-TG- 15-13 (ST-UB-NOC-15003453 dated March 12, 2015.3.2 NUREG- 1346, Techmical 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 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 fl(AI) Function in OTAT Setpoint Calculation," NF-TG-1 1-93 (ST-UB-NOC-1 1003215) dated November 10, 2011.3.7 Document RSE-U2, Rev. 5, "Unit 2 Cycle 18 Reload Safety Evaluation and Core Operating Limits Report." (CR Action 14-3642-9)

A discussion of the processes to be used to take these readings is provided in the basis for Teclmical 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 Thennal Power rami in excess of 5% of RTP per minute or a Thennal 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 limiting flow measurement uncertainty of 2.8% from Reference 3.5.

Unit 2 Cycle 18 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 8 of 16 Figure 1 Reactor Core Safety Limits -Four Loops in Operation 680 0 0 600 0 20 40 60 80 100 120 140 Rated Thermal Power (%)

lll " Unit 2 Cycle 18 Nuclear Operating Company Core Operating Limits Report Rev. 0 SN fPage 9 of 16 Figure 2 Required Shutdown Margin for Modes 3 & 4 7.0 4.0 3.0 o 2.0 0 400 800 1200 1600 2000 2400 RCS Critical Boron Concentration (ppm)(for ARI minus most reactive stuck rod) d llAIl AM Unit 2 Cycle 18 Nuclear Operating Company Core Operating Limnits Report Rev. 0 W A N-Page 10 of 16 Figure 3 Required Shutdown Margin for Mode 5 p4.0 3.0~2.0 0 400 800 1200 1 i')0 2000 RCS Critical Boron Concentration (ppm)(for ARI minus most reactive stuck rod)2400 d lATFAM Unit 2 Cycle 18 Nuclear Operating Company Core Operating Linits Report Rev. 0 Page 11 of 16 Figure 4 MTC versus Power Level 7.0 6.0 5.0 4.0 0-2 0 2.0 c.l L..1.0 0.0-1.0-2.0-3.0 0 10 20 30 40 50 60 70 80 90 100 Rated Thermal Power (%)

l& l Al Unit 2 Cycle 18 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 12 of 16 Figure 5 Control Rod Insertion Limits* versus Power Level 260 240 220 200 180 160 140 120 0 100 80 60 40 20 0 I I ( 23 259): 122 Step Overlap S(23 .258): 121Step Overlap KF- F -VV-HF~f-[I-79 259 ): 122 Step Overlap 79 258): 121 Step Overlap[, I J 1 1 1 1 1 !A 1 1- ....I ..10 1 0, 4174*~n CotoBakAialedwihrw toFlOuPoiin Fulwtdawlhllb tecndto whr htdw n IIntrol.banksnkreatthepositi n k o e I Y B UY i iOO: i 6i"'00, t 00, : :1 0 ,1 4 Full witdan shl/etecniin hr hudxv n 29 0 i coto bakaea h!oiinofi r28o 5 tp*FF~ -li I t dan T 0 10 20 30 40 50 60 70 80 90 100 Rated Thermal Power (%)

rllll~lB Unit 2 Cycle 18 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 13 of 16 Figure 6 AFD Limits versus Power Level 120 110 100 90 80 W 70 60 50 40 30 20 10 0---I i-190) --Ac- 11 ,90-) I.L_ -'3,0-tUnaccepttable!

Unacceptable i // JAcceptable

., 3~~ 1 0)L-(15 -I ' i '-50 30-20-10 0 10 20 30 40 50 Axial Flux Difference

(% Delta-I) d iPWA Unit 2 Cycle 18 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 14 of 16 Figure 7 K(Z) -Normalized FQ(Z) versus Core Height 1.2 1.1~- LL K 1.1.11 1 II 1.0 0.9 0.8! i i i I i ! ! i i i i ! I i i ! i i ! i ! i i f i i i i i i i i i i i i i i i i i 11 i i ! [ i i ! ! i i i ! ! i ! i i i i ! ! ý_ýI I I 1Thfl WIWitLflI11HIH1vIt~itIT1t4-I

[4-1171 6. 0.7 0 0.6._N 0.5 0.4 0.3 0.2 0.1 0.0.... ..iiii ii ii ..: I : .: i .,.... ,,,. -,-.-...LLLILLLCore Elev. (tt) FQ K(Z)FF -0.0 2.55 1.0 11-.L-- LJL1 I Ii ft7.0 2.55 io1.0 4 4Tý4 14.0 2.359 0.925L11 1.'-'--I -E TT rvvv~v11IThTh1~mThITh7TmTITh ITrniml7vvWI I I~F~I 0 ; I ; ' 0 ; I i i 0 1 ý I : i I I I i 0 : : I ý 0 1 1 1 1 K I I I ; 0 ý I 1 0 1 1 1 1 0 1 1 M I M M 1 1 1 i n l i l M I i _0 -1 i i i i I ! i i ! i ! ! ! I i ; ! i i ! ! ! ! i i i i i ! i I ý I i ! ýýI 0 1 2 3 4 5 6 7 8 Core Height (ft)9 10 11 12 13 14 llIr" Unit 2 Cycle 18 Nuclear Operating Company Core Operating Limits Report Rev. 0ý AF rf -Page 15 of 16 Table 1 (Part 1 of 2)Unrodded Fxy for Each Core Height for Cycle Burnups Less Than 10000 MWD/MTU Core Height (Ft.)Axial Point Unrodded Fxy Core Height (Ft.)Axial Point Unrodded Fxy 14.0 13.8 13.6 13.4 13.2 13.0 12.8 12.6 12.4 12.2 12.0 11.8 11.6 11.4 11.2 11.0 10.8 10.6 10.4 10.2 10.0 9.8 9.6 9.4 9.2 9.0 8.8 8.6 8.4 8.2 8.0 7.8 7.6 7.4 7.2 7.0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 23 24 25 26 27 28 29 30 31 32 33 34 35 36 7.337 5.734 4.131 2.763 2.576 2.267 2.184 2.180 2.113 2.048 2.012 2.031 2.082 2.067 2.004 1.960 1.949 1.937 1.914 1.931 1.987 2.001 1.930 1.899 1.895 1.886 1.886 1.905 1.974 2.040 1.973 1.910 1.897 1.892 1.884 1.872 6.8 6.6 6.4 6.2 6.0 5.8 5.6 5.4 5.2 5.0 4.8 4.6 4.4 4.2 4.0 3.8 3.6 3.4 3.2 3.0 2.8 2.6 2.4 2.2 2.0 1.8 1.6 1.4 1.2 1.0 0.8 0.6 0.4 0.2 0.0 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 1.923 1.992 1.965 1.894 1.865 1.903 1.918 1.925 1.974 2.045 2.065 2.022 1.957 1.976 1.984 1.978 1.984 2.015 2.033 1.988 1.938 1.941 1.946 1.953 1.987 2.044 2.050 1.973 1.910 1.953 2.275 3.044 4.342 6.191 9.240 dL FM Unit 2 Cycle 18 Nuclear 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 (Ft.)Axial Point Unrodded Fxy Core Height (Ft.)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 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 6.341 5.072 3.803 2.720 2.580 2.288 2.131 2.074 2.019 2.007 2.007 2.041 2.092 2.083 2.032 2.009 2.023 2.032 2.036 2.073 2.137 2.164 2.118 2.079 2.097 2.110 2.116 2.131 2.183 2.233 2.186 2.142 2.142 2.154 2.165 2.172 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 2.217 2.263 2.224 2.156 2.118 2.111 2.101 2.085 2.105 2.145 2.141 2.077 2.025 2.020 2.007 1.991 1.992 2.029 2.055 1.993 1.938 1.912 1.882 1.865 1.883 1.927 1.955 1.933 1.946 2.056 2.419 3.134 4.225 5.745 8.413