NOC-AE-14003124, Unit 1 Cycle 19 Core Operating Limits Report
ML14127A010 | |
Person / Time | |
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Site: | South Texas ![]() |
Issue date: | 04/23/2014 |
From: | Dunn R F South Texas |
To: | Document Control Desk, Office of Nuclear Reactor Regulation |
References | |
NOC-AE-14003124 | |
Download: ML14127A010 (18) | |
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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 Operating Limits Report, Revision 0 A00 (ýJm STI: 33857840 NOC-AE-1 4003124 Page 2 of 2 cc: (paper copy)(electronic copy)Regional Administrator, Region IV U. S. Nuclear Regulatory Commission 1600 East Lamar Boulevard Arlington, TX 76011-4511 Balwant K. Singal Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North (MS 8 B1)11555 Rockville Pike Rockville, MD 20852 NRC Resident Inspector U. S. Nuclear Regulatory Commission P. O. Box 289, Mail Code: MN116 Wadsworth, TX 77483 Jim Collins City of Austin Electric Utility Department 721 Barton Springs Road Austin, TX 78704 A. H. Gutterman, Esquire Morgan, Lewis & Bockius LLP Balwant K. Singal U. S. Nuclear Regulatory Commission John Ragan Chris O'Hara Jim von Suskil NRG South Texas LP Kevin Polio Cris Eugster L. D. Blaylock City Public Service Peter Nemeth Crain Caton & James, P.C.C. Mele 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 1 Cycle 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 1 Cycle 19 Nuclear Operating Company 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, 'r 2 = 3 sec T3 measured reactor vessel AT lag time constant, 3-3 = 2 sec T4 measured reactor vessel average temperature lead/lag time constant, -3 4 = 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 K 2 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%, f 1 (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, T 1 , = 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, T 7 = 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 f 2 (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 ( )R 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 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).
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 Unit 1 Cycle 19 Nuclear Operating Company 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 0 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 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.
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 RCS Critical Boron Concentration (ppm)(for ARI minus most reactive stuck rod)2400 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 RCS Critical Boron Concentration (ppm)(for ARI minus most reactive stuck rod)2400 WlIMM Unit 1 Cycle 19 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 1 Iof 16 Figure 4 MTC versus PowerLevel 7.0 6.0 5.0"4.0* 3.0-2.01.0 I0.0-1.0-2.0-3.0 Unacceptable Acceptable 4- ~ I -+ 4- t --I -+ - -4 + --f -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 260 240 220 200 180 160 c, 140 O120 100 80 60 40 20 80..........
..(79,259):..122 Stepl Overlap I l l lI I l l lIlI I .-I 1(23, 259): 122 Step Overlap (21,254):
117StepOverlap I I I I I -I I -I I I I l l l I l l l l lA g I I l l l l I I l l 00 -( 79, 259 ):122 Step Overlap-( 77,254): 117 Step Overlap f iF [1 fl Of A 00, OF wO Bank B or M 202 1,00, 1 1 100, 1-74)"0 F -T-F-F-F-T-O-e------ooe-- Bank C I I 100 ,0"00, oe oe-4+?0 iamn 00, 00, 0,65) 1 1 '00, or 4B OF= M .Alt A I* Control Bank A is already withdrawn to Full Out Position.(29,0) -Fully withdrawn shall be the condition where shutdown and IM -ý00 -control banks are at the position of either 254 or 259 steps 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 120 110 100 90 80 I I I I I I i i i i i i i i i i i-!...........I I I I I I I I I I 1~0 E 70 60 50 40 30 20 10 0 (-1 $0 11 90 Unacceptable Unacceptable
--Acceptable.I I ET-50 30-20-10 0 10 20 30 40 50 Axial Flux Difference
(% Delta-I)
A lPAlI 3A Unit 1 Cycle 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 1.0 0.9 0.8............i 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! I I I ! ! ! ! 1 1 1 1 1- ----1 1 ! II f 4 4 I---! 1 1 -!- 1 --- 4 --J' 1 1-4 1- -1 I- I -f -I -H I h WhI I I ...I ! ii i i ii i ! i i i i i i i f i ! i ! i ! i iI i: :1 I: :1 :1 , i. I , , , , , .....1. ....I ::::::::I:::::::::I
- I::::I.::
I ::I::::I::::,:.
I I :::I I 6. 0.7=0.6.t 0.5 z 0.4...........................................................
- .:I::,':................
-ore Elev. (ft) FQ K(Z)_ _ _0.0 2.55 1.0 7.0 2.55 1.0 ----_- _ _14.0 2.359 0.925 rn-mi rnmmnmmI .... I Tvwrvnvl 0.3 0.2 0.1 0.0 0 .I I I M , 1 0 1 1 0 1 1 1 M I I 1 0 , I , 0 , , ; I i I I I 1 0 1 1 i 1 0 1 1 1 ; 0 ; ; I I i I i .i 0 iý=r m=I~h tnhlI ~~ m 1 H 1 4 + 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) l IM M Unit 1 Cycle 19 Nuclear Operating Company 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 (Ft.)Axial Point Unrodded Fxy Core Height (Ft.)Axial Point Unrodded Fxv 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 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 7.641 5.953 4.255 2.796 2.535 2.331 2.208 2.123 2.056 2.004 1.971 1.974 2.003 2.005 1.973 1.941 1.920 1.900 1.888 1.902 1.938 1.957 1.930 1.905 1.909 1.913 1.912 1.929 1.990 2.051 2.004 1.938 1.912 1.888 1.856 1.833 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 1.870 1.919 1.901 1.855 1.842 1.855 1.859 1.862 1.906 1.968 1.991 1.951 1.920 1.930 1.934 1.930 1.945 1.988 2.015 1.975 1.932 1.935 1.935 1.909 1.894 1.919 1.924 1.853 1.797 1.834 2.146 2.873 4.096 5.826 8.716 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 (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 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 6.466 5.154 3.840 2.707 2.517 2.322 2.165 2.080 2.036 2.012 2.008 2.021 2.053 2.058 2.040 2.030 2.033 2.033 2.036 2.063 2.109 2.137 2.119 2.100 2.106 2.110 2.112 2.124 2.163 2.203 2.179 2.151 2.150 2.155 2.160 2.168 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 2.203 2.241 2.218 2.166 2.132 2.119 2.102 2.086 2.103 2.139 2.138 2.085 2.040 2.032 2.016 1.998 1.998 2.032 2.056 2.002 1.950 1.930 1.904 1.884 1.894 1.937 1.967 1.947 1.961 2.070 2.439 3.175 4.298 5.843 8.575