ML16333A002
| ML16333A002 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 11/17/2016 |
| From: | Dunn R South Texas |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NOC-AE-16003419, STI 34398857 | |
| Download: ML16333A002 (18) | |
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Nuclear Operating Company 11111' Ill -
South Texas Project Electric Generating Station P.O. Sox 289 Wadsworth, Texas 77483 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 19 Core Operating Limits Report November 17, 2016 NOC-AE-16003419 10 CFR 50.36 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 19. The report covers the core design changes made dudng the 2RE18 refueling outa_ge.
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) 972-77 43.
mk Roland F. Dunn
- Manager, Nuclear Fuel & Analysis
Attachment:
South Texas Project Unit 2 Cycle 19 Core Operating Limits Report, Revision O
~~
STI: 34398857
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. 0. Box 289, Mail Code: MN116 Wadsworth, TX 77 483 NOC-AE-16003419 Page 2 of 2 (electronic copy)
Morgan. Lewis & Beckius LLP Steve Frantz, Esquire U.S. Nuclear Regulatory Commission Lisa M. Regner NRG South Texas LP Chris O'Hara Jim von Suski!
Skip Zahn CPS Energy Kevin Pollo Cris Eugster L. D. Blaylock City of Austin Elaina Ball John Wester Texas Dept. of State Health Services Helen Watkins Robert Free
~....,,
Nuclear Operating Company SOUTH TEXAS PROJECT Unit 2Cycle19 CORE OPERATING LIMITS REPORT Revision 0 Core Operating Limits Repo1t Page I of 16
Unit 2Cycle19 Nuclear Operating Company Core Operating Limits Report Rev. 0 Page 2of16 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Repo1t for STPEGS Unit 2 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 Teclmical 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 MOD ERA TOR 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, pressmizer 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.
Nuclear Operating Company Unit 2Cycle19 Core Operating Limits Report Rev. 0 Page 3 of 16 2.2.2 The Over-temperature ~T and Over-power ~T setpoint parameter values are listed below:
Over-temperature ~ T Setpoint Parameter Values
'CJ measured reactor vessel ~T lead/lag time constant, 'CJ = 8 sec i:2 measured reactor vessel ~ T lead/lag time constant, i:2 = 3 sec i;3 measured reactor vessel ~ T lag time constant, i;3 = 2 sec i;4 measured reactor vessel average temperature lead/lag time constant, 1:4 = 28 sec i:s measured reactor vessel average temperature lead/lag time constant, i:s = 4 sec 1:6 measured reactor vessel average temperature lag time constant, 1:6 = 2 sec KJ Overtemperature ~ T reactor tlip setpoint, KJ = 1.14 K2 Overtemperature ~T reactor hip setpoint Tavg coefficient, K2 = 0.028/
0 P K3 Overtemperature ~ T reactor tlip setpoint pressure coefficient, K3 = 0.0014 3/psi T'
Nominal full powerTavg, T'::; 592.0 °P P'
Nominal RCS pressure, P' = 2235 psig fi(~l) 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 duling plant staiiup tests such that:
(1)
For qi - qb between-70% and +8%, f1(M) = 0, where qi and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qi+ qb is total THERMAL POWER in percent of RATED THERMAL POWER; (2)
For each percent that the magnitude of qi - qb exceeds -70%, the ~T 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 qt - qb exceeds +8%, the LiT 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 ~ T Setpoint Parameter Values
'CJ measured reactor vessel ~T lead/lag time constant, 'CJ= 8 sec i:2 measured reactor vessel ~ T lead/lag time constant, i:2 = 3 sec i;3 measured reactor vessel ~ T lag time constant, T3 = 2 sec T6 measured reactor vessel average temperature lag time constant, T6 = 2 sec
-r7 Time constant utilized in the rate-lag compensator for T avg, 1:7 = 10 sec K4 Overpower~ T reactor hip setpoint, K4 = 1.08 Ks Overpower ~T reactor hip setpoint Tavg rate/lag coefficient, Ks= 0.02/
0P for increasing average temperature, and Ks = 0 for decreasing average temperahlre K6 Overpower ~T reactor hip setpoint Tavg heah1p coefficient K6 = 0.002/
0P for T > T", and K6 = 0 for T ::; T" T"
Indicated full power Tavg, T"::; 592.0 °P fa(M) = 0 for all (~I)
Nuclear Operating Company Unit 2 Cycle 19 Core Operating Limits Report 2.3 SHUTDOWN MARGIN (Specification 3.1.1.1):
The SHUTDOWN MARGIN shall be:
2.3.l Greater than 1.3% Lip 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):
Rev. 0 Page4of16 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% RA TED THERMAL POWER),
HFP vessel average temperah1re is 592 °F.
2.4.4 The Revised Predicted near-EOL 300 ppm MTC shall be calculated using the algoritlun from Technical Specification 6.9.1.6.b. l 0:
Revised Predicted MTC = Predicted MTC + AFD C01Tection - 3 pcm/°F If the Revised Predicted MTC is less negative than the COLR Section 2.4.3 limit and all of the benclunark data contained in the surveillance procedure are met, then an MTC measurement in accordance with S.R. 4.1.1 Jb 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).
Nu cl.ear Operating Company Unit 2 Cycle 19 Core Operating Limits Report Rev. 0 Page 5of16 2.6 AXIAL FLUX DIFFERENCE (Specification 3.2.1):
2.6.l 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 ACCEPT ABLE OPERATION p01iion 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 (F~lP) within specific core planes shall be:
2.7.3.l 2.7.3.2 2.7.3.3 Less than or equal to 2.102 for all cycle bumups for all core planes containing Bank "D" control rods, and Less than or equal to the appropriate core height-dependent value from Table 1 for all unrodded core planes.
PFxy= 0.2.
These Fxy limits were used to confinn 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 vaiiations in the axial xenon and power distributions, as desclibed 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 cliteria of 10 CFR 50.46.
2.7.4 Core Power Distribution Measurement Uncertainty for the Heat Flux Hot Cha1111el Factor 2.7.4.1 If the Power Distribution Monit01ing 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/lOO))*UE Where:
UQ =
Unce1iainty for power peaking factor as defined in Equation 5-19 from the document referenced by Technical Specification 6.9.1.6.b.ll U E =
Engineering uncertainty factor of 1. 03.
This unce1iainty is calculated and applied automatically by the Power Distiibution Monitoring System (PDMS).
Nuclear Operating Company
'2.7.4.2 Unit 2 Cycle 19 Core Operating Limits Report Rev. Q_,
Page 6of16 If the moveable detector system is used, the core power distribution measurement tmcertainty (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.
U E =
Engineering unce1iainty factor of 1. 03.
2.8 ENTHALPY RISE HOT CHANNEL FACTOR (Specification 3.2.3):
2.8.l FfilJP = 1.62 2.8.2 PF1rn = 0.3 2.8.3 Core Power Distribution Measurement Uncertainty for the Enthalpy Rise Hot Channel Factor 2.8.3.l 2.8.3.2 If the Power Distribution Monitming System (PDMS) is operable, as defined in the Technical Requirements Manual Section 3.3.3.12, the core power distribution measurement uncertainty (UFt.H) to be applied to the F~H using the PDMS shall be the greater of:
UFL\\H = 1.04 OR UFL\\H = 1.0 + (Ut.H/100)
Where:
Ut.tt = Uncertainty for power peaking factor as defined in Equation 5-19 from the document referenced in Teclmical Specification 6.9. l.6.b.11.
This unce1iainty is calculated and applied automatically by the Power Distribution Monitoring System.
If the moveable detector system is used, the core power distribution measurement unce1iainty (UFt.H) shall be:
UFL\\H = 1.04
Nuclear Operating Company Unit 2Cycle19 Core Operating Limits Report Rev. 0 Page 7of16 2.9 DNB PARAMETERS (Specification 3.2.5):
2.9.1 The following DNB-related parameters shall be maintained within the following limits (nominal values from Reference 3.1, as annotated below): 1 2.9.1.1 2.9.1.2 2.9.1.3 Reactor Coolant System Tavg ::; 595 °F 2, Pressurizer Pressure > 2200 psig 3, Minimum Measured Reactor Coolant System Flow > 403,000 gpm4.
3.0 REFERENCES
2 4
3.1 Letter from J.M. Ralston (Westinghouse) to R. F. Dunn (STPNOC), "South Texas Project Electric Generating Station Unit 2 Cycle 19 Final Reload Evaluation" NF-TG-16-49 (ST-UB-NOC-16003543) dated August 4, 2016.
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. 6, "Loop Uncertainty Calculation for Narrow Range Pressurizer Pressure Monitoring h1strumentation," 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 f1(Lil) Function in OTLiT Setpoint Calculation," NF-TG-11-93 (ST-UB-NOC-11003215) dated November 10, 2011.
3.7 Document RSE-U2, Rev. 6, "Unit 2 Cycle 19 Reload Safety Evaluation and Core Operating Limits Rep01t." (CR Action 15-11357-9)
A discussion of the processes to be used to take these readings is provided in the basis for Technical Specification 3.2.5.
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 The1mal Power step in excess of 10% RTP. Per Technical Specification 3.2.5 Bases, this includes a 10.7 psi measmement uncertainty as read on the QDPS display, which is bounded by the 9.6 psi averaged measurement calculated in Reference 3.4.
Includes the most limiting flow measurement uncertainty of 2.8% from Reference 3.5.
Nuclear Operating Company Unit 2 Cycle 19 Core Operating Limits Report Figure 1 680 Reactor Core Safety Limits - Four Loops in Operation 66 IC 2, 664.52 )I Unacceptable i
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Nuclear Operating Company Rev. 0 Page 12of16 Figures Control Rod Insertion Limits* ve!'sus Power Level 260
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Nuclear Operating Company Unit 2 Cycle 19 Core Operating Limits Report Rev.O Page 14of16 Figure 7 K(Z)- Nommlized FQ(Z) versus Core Height
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Unit 2 Cycle 19 Nuclear.Operating Company Core Operating Lim.its Report Rev. 0 Page 15of16 Table 1 (Part 1 of 2)
Unrodded Fxy for Each Core Height for Cycle Burnups Less Than 9000 MWD/MTU Core Height Axial Unrodded Core Height Axial Unrodded (Ft.)
Point Fxy (Ft.)
Point Fxy 14.0 1
7.705 6.8 37 2.004 13.8 2
6.000 6.6 38 2.039 13.6 3
4.305 6.4 39 2.014 13.4 4
2.862 6.2 40 1.969 13.2 5
2.639 6.0 41 1.946 13.0 6
2.319 5.8 42 1.947 12.8 7
2.196 5.6 43 1.949 12.6 8
2.145 5.4 44 1.946 12.4 9
2.077 5.2 45 1.976 12.2 10 2.015 5.0 46 2.022 12.0 11 1.978 4.8 47 2.044 11.8 12 1.993 4.6 48 1.995 11.6 13 2.043 4.4 49 1:939 11.4 14 2.029 4.2 50 1.950 11.2 15 1.969 4.0 51 1.959 11.0 16 1.928 3.8 52 1.952 10.8 17 1.918 3.6 53 1.963 10.6 18 1.896 3.4 54 2.008 10.4 19 1.877 3.2 55 2.044 10.2 20 1.893 3.0 56 1.995 10.0 21 1.940 2.8 57 1.942 9.8 22 1.958 2.6 58 1.945 9.6 23 1.913 2.4 59 1.950 9.4 24 1.880 2.2 60 1.957 9.2 25 1.904 2.0 61 1.989 9.0 26 1.923 1.8 62 2.048 8.8 27 1.937 1.6 63 2.062 8.6 28 1.963 1.4 64 1.996 8.4 29 2.019 1.2 65 1.951 8.2 30 2.063 1.0 66 2.019 8.0 31 2.036 0.8 67 2.368 7.8 32 1.998 0.6 68 3.180 7.6 33 1.992 0.4 69 4.538 7.4 34 1.989 0.2 70 6.468 7.2 35 1.982 0.0 71 9.669 7.0 36 1.970
Unit 2Cycle19 Nuclear Operating Company
.... Core-Operating Limits Report Rev.O Page 16of16 Table 1 (Part 2 of 2)
Unrodded Fxy for Each Core Height for Cycle Burnups Greater Than or Equal to 9000 MWD/MTU Core Height Axial Umodded Core Height Axial Umodded (Ft.)
Point Fxy (Ft.)
Point Fxy 14.00 1
6.451 6.80 37 2.208 13.80 2
5.157 6.60 38 2.253 13.60 3
3.863 6.40 39 2.215 13.40 4
2.756 6.20 40 2.149 13.20 5
2.605 6.00 41 2.113 13.00 6
2.321 5.80 42 2.105 12.80 7
2.161 5.60 43 2.094 12.60 8
2.101 5.40 44 2.078 12.40 9
2.044 5.20 45 2.099 12.20 10 2.006 5.00 46 2.139 12.00 11 2.008 4.80 47 2.136 11.80 12 2.042 4.60 48 2.074 11.60 13 2.096 4.40 49 2.023 11.40 14 2.092 4.20 50 2.018 11.20 15 2.047 4.00 51 2.005 11.00 16 2.001 3.80 52 1.989 10.80 17 2.032 3.60 53 1.990 10.60 18 2.041 3.40 54 2.027 10.40 19 2.042 3.20 55 2.053 10.20 20 2.075 3.00 56 1.993 10.00 21 2.134 2.80 57 1.931 9.80 22 2.161 2.60 58 1.903 9.60 23 2.118 2.40 59 1.886 9.40
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