ML15062A275

Wolf Creek Generating Station - Pressure and Temperature Limits Report, Revision 2

ML15062A275
Person / Time
Site:	Wolf Creek
Issue date:	02/24/2015
From:	Koenig S R Wolf Creek
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA 15-0015
Download: ML15062A275 (27)

Text

WOLF CREEKNUCLEAR OPERATING CORPORATIONSteven R. KoenigManager Regulatory AffairsFebruary 24, 2015RA 15-0015U. S. Nuclear Regulatory CommissionATTN: Document Control DeskWashington, DC 20555

Subject:

Docket No. 50-482: Wolf Creek Generating Station Pressure andTemperature Limits Report, Revision 2Gentlemen:Enclosed is Revision 2 of the Wolf Creek Generating Station (WCGS) Pressure andTemperature Limits Report (PTLR). Revision 2 of the PTLR is being submitted pursuant toSection 5.6.6, "Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITSREPORT (PTLR)," of the WCGS Technical Specifications.This letter contains no commitments. If you have any questions concerning this matter, pleasecontact me at (620) 364-4041.Sincerely,Steven R. Koen'SRK/rltEnclosurecc: M. L. Dapas (NRC), w/eC. F. Lyon (NRC), w/eN. F. O'Keefe (NRC), w/eSenior Resident Inspector (NRC), w/eP.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831An Equal Opportunity Employer M/F/HCNET Enclosure to RA 15-0015WOLF CREEK GENERATING STATION -UNIT IPRESSURE AND TEMPERATURE LIMITS REPORT, Revision 2(25 pages)

WOLF CREEK GENERATING STATION -UNIT IPRESSURE AND TEMPERATURE LIMITS REPORTRevision 2 PRESSURE AND TEMPERATURE LIMITS REPORTTable of ContentsSection Page1.0 Reactor Coolant System (RCS) PRESSURE AND 1TEMPERATURE LIMITS REPORT (PTLR)2.0 Operating Limits 12.1 RCS Pressure and Temperature Limits 12.2 Low Temperature Overpressure Protection System 13.0 Reactor Vessel Material Surveillance Program 94.0 Reactor Vessel Surveillance Data Credibility 95.0 Supplemental Data Tables 146.0 References 15Wolf Creek -Unit 1Revision 2 PRESSURE AND TEMPERATURE LIMITS REPORTList of FiguresFigure Page2.1-1 Wolf Creek Reactor Coolant System Heatup Limitations (Heatup 3Rates of 60 and 1 00°F/hr). Applicable to 54 EFPY (WithoutMargins for Instrumentation Uncertainty)2.1-2 Wolf Creek Reactor Coolant System Cooldown Limitations 5(Cooldown Rates of 0, 20, 40, 60 and 1 OO0F/hr) Applicable to 54EFPY (Without Margins for Instrumentation Uncertainty)2.2-1 Maximum Allowed PORV Setpoint for the Low Temperature 7Overpressure Protection SystemList of Tables2.1-1 Wolf Creek Heatup Data at 54 EFPY Without Margins for 4Instrumentation Uncertainty2.1-2 Wolf Creek Cooldown Data at 54 EFPY Without Margins for 6Instrumentation Uncertainty2.2-1 Data Points for Maximum Allowed PORV Setpoint 8Wolf Creek -Unit IiiRevision 2 PRESSURE AND TEMPERATURE LIMITS REPORT1.0 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)This PTLR for WCGS has been prepared in accordance with the requirements of TechnicalSpecification (TS) 5.6.6. The TS addressed in this report are listed below:LCO 3.4.3 RCS Pressure and Temperature (P/T) LimitsLCO 3.4.12 Low Temperature Overpressure Protection (LTOP) Systems2.0 Operating LimitsThe parameter limits for the specifications listed in Section 1.0 are presented in the followingsubsections. The limits were developed using a methodology that is in accordance with theNRC-approved methodology specified in Specification 5.6.6 (Ref. 1). In addition, the newWolf Creek heatup and cooldown P/T limit curves were developed using ASME Code CaseN-641, which allows the use of the static crack initiation fracture toughness curve (Kjc).NRC approval of this methodology was received in Reference 2. NRC acceptance forreferencing this methodology was received in Amendment No. 180 (Ref. 3).The revised P/T Limit curves account for a requirement of 10 CFR 50, Appendix G, that thetemperature of the closure head flange and vessel flange regions must be at least 120OFhigher than the limiting RTNDT for these regions when the pressure exceeds 20% of thepreservice hydrostatic test pressure (3106 psig).2.1 RCS Pressure and Temperature (P/T) Limits (LCO 3.4.3)2.1.1 The RCS temperature rate-of-change limits are (Ref. 2)a. A maximum heatup of 100°F in any 1-hour period.b. A maximum cooldown of 1000F in any 1-hour period.c. A maximum temperature change of 10°F in any 1-hour period duringinservice hydrostatic and leak testing operations above the heatup andcooldown limit curves.2.1.2 The RCS P/T limits for heatup, cooldown, inservice hydrostatic and leaktesting, and criticality are specified by Figures 2.1-1 and 2.1-2 (Ref. 5).2.2 Low Temperature Overpressure Protection (LTOP) System Setpoints (LCO 3.4.12)The power-operated relief valves (PORVs) shall each have lift settings inaccordance with Figure 2.2-1. The LTOP System (Cold Overpressure MitigationSystem/PORVs) arming temperature is 3680F. These lift setpoints have beendeveloped using the NRC approved methodologies specified in TechnicalSpecification 5.6.6.Wolf Creek -Unit 11Revision 2 PRESSURE AND TEMPERATURE LIMITS REPORT2.2 (continued)The revised Wolf Creek heatup and cooldown P/T limit curves (Ref. 5) weregenerated by Westinghouse as part of the testing and analysis of pressure vesselsurveillance samples from Capsule X of the WCGS reactor pressure vessel radiationsurveillance program. It should be noted that the static crack initiation fracturetoughness curve (Ki,), as given in ASME Code Case N-641 and included in TopicalReport WCAP-1 4040-A, Revision 4 (Ref. 4) as an option for the development of P/Tlimit curves, is used as a basis for developing P/T limit curves. The NRC staffaccepts this Code Case as an option for the development of P/T limit curves, as theuse of optional guidelines for the development of P/T limit curves also meets theregulatory requirements of Appendix G to 10 CFR Part 50 and the guidanceprovided in SRP Section 5.3.2.However, the use of Code Case N-641 presently includes a restriction on themaximum allowed PORV setpoint for the LTOP system, which is derived based onthe revised heatup and cooldown limit curves. The maximum pressure for the LTOPis 100% of the pressure allowed by the P/T limit curves. This is different from theprevious analysis that used the Kia (dynamic crack initiation/crack arrest) fracturetoughness curve, along with the use of ASME Code Case N-514, which allows a10% relaxation of the Appendix G limits below the LTOP enabling temperature.As a result, the revised PORV setpoint limits for the LTOP system are determinedbased on 100% of the pressure allowed by the revised P/T limit curves, and theanalysis results of the limiting design basis mass and heat input transients. Thethermal hydraulic analysis for the mass and heat input transients use the samespecialized version of the LOFTRAN code, previously approved by the NRC staff forthis type of application.Operation with a PORV setpoint less than or equal to the maximum setpoint ensuresthat Appendix G criteria will not be violated with consideration for: (1) process andinstrumentation uncertainties; (2) single failure. To ensure mass and heat inputtransients more severe than those assumed cannot occur, it is required to lockoutboth Safety Injection pumps and one centrifugal charging pump (one centrifugalcharging pump and the normal charging pump are operational) while in MODES 4, 5,and 6 with the reactor vessel head installed, and limit the heat input due to starting areactor coolant pump, if secondary temperature is more than 50°F above reactorcoolant temperature.Wolf Creek -Unit 12Revision 2 PRESSURE AND TEMPERATURE LIMITS REPORTMATERIAL PROPERTY BASISLIMITING MATERIAL: LOWER SHELL PLATE R2508-3LIMITING ART VALUES AT 54 EFPY: 1/4T, 1040F3/4T, 930F25002250Leak Test Limit2000Unacceptable Acceptable1750 Operation A t0 OperationC. 1500 Heatup Rate ._,_.60 Deg. F/Hr'- Critical Limit-1 250 Heatup Rate __60 Deg. F/Hr0100 Deg. F/HrO .._ .C ditica l L im it". 1000 -H -----00ODeg. F/Hr750,*500.*-Boltup Criticality Limit based on, Temnp. inservice hyd rostatic test250 -60OF temperature (164°F) for the-service period up to 54 EFPY0The lower limit for RCSpressure is -14.7 psig0 50 100 150 200 250 300 350 400 450 500 550Moderator Temperature (Deg. F)FIGURE 2.1-1 Wolf Creek Reactor Coolant System Heatup Limitations (Heatup Rates of 60and 100°OF/hr) Applicable for the First 54 EFPY (Without Margins forInstrumentation Uncertainty)Wolf Creek -Unit 13Revision 2 PRESSURE AND TEMPERATURE LIMITS REPORTTABLE 2.1-1Wolf Creek Heatup Limits at 54 EFPYWithout Margins for Instrumentation Uncertainty60°Flhr 60°Flhr Crit. 100°Flhr 100°Flhr Crit. Leak Test LimitLimit LimitTemp. Press. Temp. Press. Temp. Press. Temp. Press. Temp. Press.(OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig)60 -14.7 164 -14.7 60 -14.7 164 -14.7 147 200060 621 164 621 60 621 164 621 164 248565 621 164 621 65 621 164 62170 621 164 621 70 621 164 62175 621 164 621 75 621 164 62180 621 164 621 80 621 164 62185 621 164 621 85 621 164 62190 621 164 621 90 621 164 62195 621 164 621 95 621 164 621100 621 164 621 100 621 164 621105 621 164 621 105 621 164 621110 621 164 621 110 621 164 621115 621 164 621 115 621 164 621120 621 165 621 120 621 165 621125 621 170 621 125 621 170 621130 621 175 621 130 621 175 621135 621 180 621 135 621 180 621140 621 180 1074 140 621 180 881140 1074 185 1129 140 881 185 915145 1129 190 1190 145 915 190 953150 1190 195 1257 150 953 195 995155 1257 200 1333 155 995 200 1043160 1333 205 1416 160 1043 205 1096165 1416 210 1507 165 1096 210 1156170 1507 215 1609 170 1156 215 1222175 1609 220 1721 175 1222 220 1295180 1721 225 1846 180 1295 225 1376185 1846 230 1983 185 1376 230 1466190 1983 235 2134 190 1466 235 1565195 2134 240 2301 195 1565 240 1675200 2301 200 1675 245 1796205 1796 250 1930210 1930 255 2078215 2078 260 2241220 2241 265 2421225 2421IWolf Creek -Unit 14Revision 2 PRESSURE AND TEMPERATURE LIMITS REPORTMATERIAL PROPERTY BASISLIMITING MATERIAL: LOWER SHELL PLATE R2508-3LIMITING ART VALUES AT 54 EFPY: 1/4T, 1040F3/4T, 930FV)U)I-M.2500225020001750150012501000750500UnacceptableOperationAcceptableOperationCooldownRates, °F/Hrsteady-state-20-40-60-100BoltupTemperature, 607FThe lower limit for RCSpressure is -14.7 psig-. .- I ....- Ir -...1 , 11 1 pl l f l l l25000 50 100 150 200 250300 350 400 450 500 550Moderator Temperature (Deg. F)FIGURE 2.1-2Wolf Creek Reactor Coolant System Cooldown Limitations (Cooldown Rates upto 1 00°F/hr) Applicable for the First 54 EFPY (Without Margins forInstrumentation Uncertainty)Wolf Creek -Unit 15Revision 2 PRESSURE AND TEMPERATURE LIMITS REPORTTABLE 2.1-2Wolf Creek Cooldown Limits at 54 EFPYWithout Margins for Instrumentation UncertaintySteady State 200F/hr 40°Flhr 60°F/hr 100°F/hrTemp. Press. Temp. Press. Temp. Press. Temp. Press. Temp. Press.(OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig)60 -14.7 60 -14.7 60 -14.7 60 -14.7 60 -14.760 621 60 621 60 621 60 621 60 58865 621 65 621 65 621 65 621 65 61470 621 70 621 70 621 70 621 70 62175 621 75 621 75 621 75 621 75 62180 621 80 621 80 621 80 621 80 62185 621 85 621 85 621 85 621 85 62190 621 90 621 90 621 90 621 90 62195 621 95 621 95 621 95 621 95 621100 621 100 621 100 621 100 621 100 621105 621 105 621 105 621 105 621 105 621110 621 110 621 110 621 110 621 110 621115 621 115 621 115 621 115 621 115 621120 621 120 621 120 621 120 621 120 621125 621 125 621 125 621 125 621 125 621130 621 130 621 130 621 130 621 130 621135 621 135 621 135 621 135 621 135 621140 621 140 621 140 621 140 621 140 621140 1387 140 1387 140 1387 140 1387 140 1387145 1469 145 1469 145 1469 145 1469 145 1469150 1560 150 1560 150 1560 150 1560 150 1560155 1660 155 1660 155 1660 155 1660 155 1660160 1771 160 1771 160 1771 160 1771 160 1771165 1893 165 1893 165 1893 165 1893 165 1893170 2028 170 2028 170 2028 170 2028 170 2028175 2178 175 2178 175 2178 175 2178 175 2178180 2343 180 2343 180 2343 180 2343 180 2343IWolf Creek -Unit 16Revision 2 PRESSURE AND TEMPERATURE LIMITS REPORT2500-2000 !1500 --PORV Setpoint LimitsZO - PORV #1 Breakpointsc-a -M PORV #2 Breakpointso 1000500 -__ 0 50 100 150 200 250 300 350 400 450MEASURED RTD TEMPERATURE (DEG F)2 RCPs running below 100 OF4 RCPs running above 100 OFFIGURE 2.2-1Maximum Allowed PORV Setpoint for the Low Temperature OverpressureProtection SystemWolf Creek -Unit I7Revision 2 PRESSURE AND TEMPERATURE LIMITS REPORTTABLE 2.2-1Maximum Allowed PORV Setpoints/BreakpointsTemperature Pressure (psig)(OF) Max. Allowed PORV #1 PORV #260 425 415 42578 425 415 42588 425 -- --118 425 415 425158 425 415 425168 529 -- -208 529 460 525218 540 -- --268 650 570 650318 800 680 800343 910 -- --368 1127 680 800418 2350 -- --425 -- 2350 2350Wolf Creek -Unit 18Revision 2 PRESSURE AND TEMPERATURE LIMITS REPORT3.0 Reactor Vessel Material Surveillance ProgramThe reactor vessel material surveillance program is in compliance with Appendix H to 10CFR 50, entitled "Reactor Vessel Material Surveillance Program Requirements" and Section5.3 of the WCGS Updated Safety Analysis Report. The withdrawal schedule is presented inUSAR Table 5.3-11. The surveillance capsule reports are as follows:1. WCAP-1 1553, August 1987, "Analysis of Capsule U from the Wolf Creek NuclearOperating Corporation, Wolf Creek Reactor Vessel Radiation SurveillancePrograms."2. WCAP-1 3365, Revision 1, April 1993, "Analysis of Capsule Y from the Wolf CreekNuclear Operating Corporation, Wolf Creek Reactor Vessel Radiation SurveillancePrograms."3. WCAP-1 5078, Revision 1, August 1998, "Analysis of Capsule Vfrom the Wolf CreekNuclear Operating Corporation, Wolf Creek Reactor Vessel Radiation SurveillancePrograms."4. WCAP-1 6028, Revision 0, March 2003, "Analysis of Capsule X from the Wolf CreekNuclear Operating Corporation, Wolf Creek Reactor Vessel Radiation SurveillancePrograms."Note: The last W and Z capsules were withdrawn during Refueling Outage 14 (Spring2005) and are now stored in the spent fuel pool and will be retained for the extendedlicensed operating period. During Refueling Outage 14, Ex-Vessel Dosimetry wasinstalled at Wolf Creek to provide continuous monitoring of the beltline region of thereactor vessel.4.0 Reactor Vessel Surveillance Data CredibilityRegulatory Guide 1.99, Revision 2, describes general procedures acceptable to the NRCstaff for calculating the effects of neutron radiation embrittlement of the low-alloy steelscurrently used for light-water-cooled reactor vessels. Position C.2 of Regulatory Guide 1.99,Revision 2, describes the method for calculating the adjusted reference temperature andCharpy upper-shelf energy of reactor vessel beltline materials using surveillance capsuledata. The methods of Position C.2 can only be applied when two or more crediblesurveillance data sets become available from the reactor in question.To date there has been four surveillance capsules removed from the Wolf Creek reactorvessel and tested. To use these surveillance data sets, they must be shown to be credible.In accordance with the discussion of Regulatory Guide 1.99, Revision 2, there are fiverequirements that must be met for the surveillance data to be judged credible.The purpose of this evaluation is to apply the credibility requirements of Regulatory Guide1.99, Revision 2, to the Wolf Creek reactor vessel surveillance data and determine if theWolf Creek surveillance data is credible.Wolf Creek -Unit 19Revision 2 PRESSURE AND TEMPERATURE LIMITS REPORTCriterion 1: Materials in the capsules should be those judged most likely to becontrolling with regard to radiation embrittlement.The beltline region of the reactor vessel is defined in Appendix G to 10 CFR Part 50,"Fracture Toughness Requirements," as follows:"the reactor vessel (shell material including welds, heat affected zones, and plates orforgings) that directly surrounds the effective height of the active core and adjacentregions of the reactor vessel that are predicted to experience sufficient neutronradiation damage to be considered in the selection of the most limiting material withregard to radiation damage."The Wolf Creek reactor vessel consists of the following beltline region materials:* Intermediate shell plate R2005-1, 2, 3* Lower shell plate R2508-1, 2, 3" Intermediate & Lower Shell Longitudinal Weld Seams (Heat # 90146),* Intermediate & Lower Shell Circumferential Weld Seams (Heat # 90146),Per WCAP-10015, the Wolf Creek surveillance program was based on ASTM E185-79.When the surveillance program material was selected it was believed that copper andphosphorus were the elements most important to embrittlement of reactor vessel steels.Lower shell plate R2508-3 had the highest initial RTNDT and the lowest initial USE of all platematerials in the beltline region. In addition, lower shell plate R2508-3 had approximately thesame copper and phosphorous content as the other beltline plate materials. Therefore,based on the highest initial RTNDT and lowest initial upper shelf energy, lower shell plate waschosen for the surveillance program.The weld material in the Wolf Creek surveillance program was made of the same wire as allthe reactor vessel beltline welds, thus it was chosen as the surveillance weld material.Hence, this criterion is met for the Wolf Creek reactor vessel.Criterion 2: Scatter in the plots of Charpy energy versus temperature for theirradiated and unirradiated conditions should be small enough to permitthe determination of the 30 ft-lb temperature and upper shelf energyunambiguously.Plots of Charpy energy versus temperature for the unirradiated and irradiated condition arepresented in WCAP-1 6028, Revision 0, March 2003, "Analysis of Capsule X from the WolfCreek Nuclear Operating Corporation, Wolf Creek Reactor Vessel Radiation SurveillanceProgram," (Ref. 7).The scatter in the data presented in these plots is small enough to permit the determinationof the 30 ft-lb temperature and the upper shelf energy of the Wolf Creek surveillancematerials unambiguously. Hence, the Wolf Creek surveillance program meets this criterion.Wolf Creek -Unit 110Revision 2 PRESSURE AND TEMPERATURE LIMITS REPORTCriterion 3:When there are two or more sets of surveillance data from one reactor, thescatter of ARTNDT values about a best-fit line drawn as described in RegulatoryPosition 2.1 normally should be less than 28°F for welds and 170F for basemetal. Even if the fluence range is large (two or more orders of magnitude), thescatter should not exceed twice those values. Even if the data fail this criterionfor use in shift calculations, they may be credible for determining decrease inupper shelf energy if the upper shelf can be clearly determined, following thedefinition given in ASTM E185-82.The functional form of the least squares method as described in Regulatory Position 2.1 will beutilized to determine a best-fit line for this data and to determine if the scatter of the ARTNDT valuesabout this line is less than 280F for welds and less than 170F for the plate.Following is the calculation of the best-fit line as described in Regulatory Position 2.1 of RegulatoryGuide 1.99, Revision 2. In addition, the recommended NRC methods for determining credibilitywillbe followed. The NRC methods were presented to the industry at a meeting held by the NRC onFebruary 12 and 13, 1998. At this meeting the NRC presented five cases. Of the five cases, Case1 ("Surveillance data available from plant but no other source") most closely represents the situationlisted above for the Wolf Creek surveillance weld metal. Note, for the plate materials, the straightforward method of Regulatory Guide 1.99, Revision 2 will be followed.Wolf Creek -Unit 111Revision 2 PRESSURE AND TEMPERATURE LIMITS REPORTTable 4.0-1Calculation of Chemistry Factors using Wolf Creek Surveillance Capsule DataMaterial Capsule Capsule FF(b) ARTNDT(c) FF*ARTNDT FF2f(a)Lower Shell U 0.316 0.684 36.46 24.94 0.468Plate R2508-3 Y 1.19 1.05 16.03 16.83 1.10(Longitudinal) V 2.22 1.22 52.03 63.48 1.49X 3.49 1.33 61.06 81.21 1.77Lower Shell U 0.316 0.684 23.79 16.27 0.468Plate R2508-3 Y 1.19 1.05 35.39 37.16 1.10(Transverse) V 2.22 1.22 54.53 66.53 1.49X 3.49 1.33 53.96 71.77 1.77SUM: 378.19 9.656CFR2508-3 = X(FF

• ARTNDT) + X( FF2) = (378.19) + (9.656) = 39.1-FSurveillance U 0.316 0.684 27.21 18.612 0.468WeldMaterial(d) Y 1.19 1.05 45.09 47.34 1.10V 2.22 1.22 46.33 56.49 1.49X 3.49 1.33 68.36 90.92 1.77SUM: 213.362 4.828CF Surv. Weld = YI(FF

• ARTNDT) + E( FF2) = (213.362) -(4.828) = 44.1°FNotes:(a) f = Calculated Fluence (1019 n/cm2, E > 1.0 MeV). These values were re-evaluated as partof the capsule X analysis. (See Appendix D of WCAP-1 6028, Revision 0)(b) FF = fluence factor = f(c) ARTNDT values are the measured 30 ft-lb shift values taken from WCAP-1 6028, [OF].The scatter of ARTNDT values about the functional form of a best fit line drawn as described inRegulatory Position 2.1 is present in Table 4.0-2.Wolf Creek -Unit 112Revision 2 PRESSURE AND TEMPERATURE LIMITS REPORTTable 4.0-2Wolf Creek Surveillance Capsule Data Scatter about the Best-Fit Line for SurveillanceForgin MaterialsMaterial Capsule CF FF Measured Predicted Scatter <170F (BaseARTNDT(a) ARTNDT(b) ARTNDT Metals)(°F) <170F (Weld)Lower Shell U 39.1 0.684 36.46 26.74 9.72 YesPlate Y 39.1 1.05 16.03 41.06 -25.03 NoR2508-3 V 39.1 1.22 52.03 47.70 4.33 Yes(Longitudinal) X 39.1 1.33 61.06 52.00 9.06 YesLower Shell U 39.1 0.684 23.79 26.74 -2.95 YesPlate Y 39.1 1.05 35.39 41.06 -5.67 YesR2508-3 V 39.1 1.22 54.53 47.70 6.83 Yes(Transverse) X 39.1 1.33 53.96 52.00 1.96 YesSurveillance U 44.1 0.684 27.21 30.16 -2.95 YesProgram Y 44.1 1.05 45.09 46.31 -1.22 YesWeld Metal V 44.1 1.22 46.33 53.80 -7.47 YesX 44.1 1.33 68.36 58.65 9.71 YesNotes:(a) Based on measured Charpy data plotted with CVGRAPH 4.1.(b) Best estimate ARTNDT = CF

• FF, where the CF is based on the measured surveillance data.Table 4.0-2 indicates that only one data point falls outside the +/- 1 a of 1 70F scatter band for thelower shell plate R2508-3 surveillance data. One out of 8 data points is still considered credible. Noweld data point fall outside the +/- 1 a of 28 scatter band for the surveillance weld data, therefore theweld data is also credible per the third criterion.Criterion 4:The irradiation temperature of the Charpy specimens in the capsule shouldmatch the vessel wall temperature at the cladding/base metal interface within+/- 250F.The capsule specimens are located in the reactor between the neutron pads and the vessel wall andare positioned opposite the center of the core. The test capsules are in baskets attached to theneutron pads. The location of the specimens with respect to the reactor vessel beltline providesassurance that the reactor vessel wall and the specimens experience equivalent operatingconditions such that the temperatures will not differ by more than 250F. Hence this criteria is met.Wolf Creek -Unit 113Revision 2 PRESSURE AND TEMPERATURE LIMITS REPORTCriterion 5:The surveillance data for the correlation monitor material in the capsuleshould fall within the scatter band of the data base for that material.The Wolf Creek surveillance program does not contain correlation monitor material.Therefore, this criterion is not applicable to the Wolf Creek surveillance program.Conclusion:Based on the proceding responses to all five criteria of Regulatory Guide 1.99, Revision 2,Section B and 10 CFR 50.61, the Wolf Creek surveillance plate and weld data is credible.5.0 Supplemental Data TablesTable 5.0-1Table 5.0-2Table 5.0-3Table 5.0-4Table 5.0-5Table 5.0-6Table 5.0-7Comparison of Wolf Creek Surveillance Material 30-ft-lb TransitionTemperature Shifts and Upper Shelf Energy Decreases with RegulatoryGuide 1.99, Revision 2, PredictionsCalculation of Chemical Factors Using Surveillance Capsule DataWolf Creek Reactor Vessel Beltline Unirradiated Material PropertiesSummary of the Peak Pressure Vessel Neutron Fluence Values at54 EFPY used for the Calculation of Adjusted Reference Temperature(ART) ValuesSummary of Adjusted Reference Temperatures (ARTs) for the ReactorVessel Beltline Materials at the 1/4-T and %-T Locations for 54 EFPYCalculation of the Adjusted Reference Temperatures (ARTs) at 54 EFPYfor the Limiting Reactor Vessel Material (Lower Shell Plate R-2508-3)RTPTS Calculation for Wolf Creek Beltline Region Material at LifeExtension (54 EFPY)Wolf Creek -Unit 114Revision 2 PRESSURE AND TEMPERATURE LIMITS REPORT6.0 References1. Technical Specification 5.6.6, "Reactor Coolant System (RCS) PRESSURE ANDTEMPERATURE LIMITS REPORT (PTLR).-2. NRC letter dated February 27, 2004, Final Safety Evaluation for Topical ReportWCAP-14040, Revision 3, "Methodology Used to Develop Cold OverpressureMitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004.3. License Amendment No. 180, dated January 27, 2009, from Balwant K. Singal,USNRC, to Rick A. Muench, WCNOC.4 WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold OverpressureMitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004.5. WCAP-1 6029, Revision 0, "Wolf Creek Heatup and Cooldown Limit Curves for NormalOperation," May 2003.6. WCAP-1 6030, Revision 0, "Evaluation of Pressurized Thermal Shock for Wolf Creek,"May 2003.7. WCAP-16028, Revision 0, "Analysis of Capsule X from the Wolf Creek NuclearOperating Corporation, Wolf Creek Reactor Vessel Radiation SurveillanceProgram," March 2003.8. WCAP-1 5080, Revision 1, "Evaluation of Pressurized Thermal Shock for WolfCreek," September 1998.Wolf Creek -Unit I15Revision 2 PRESSURE AND TEMPERATURE LIMITS REPORTTable 5.0-1Comparison of Wolf Creek Surveillance Material 30 ft-lb Transition Temperature Shiftsand Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, PredictionsFluence(d) 30 ft-lb Transition Upper Shelf EnergyMaterials Capsule (x 1019 Temperature Shift Decreasen/cm2, Predicted Measured Predicted MeasuredE>1.0 MeV) (OF) (a) (OF) (b) (%) (a) (%) (c)Lower Shell Plate U 0.316 34.88 36.46 14.5 2R2508-3 Y 1.19 53.55 16.03 20 11V 2.22 62.22 52.03 23 13(Longitudinal) X 3.49 67.83 61.06 25 4Lower Shell Plate U 0.316 34.88 23.79 14.5 0R2508-3 Y 1.19 53.55 35.39 20 0V 2.22 62.22 54.53 23 6(Transverse) X 3.49 67.83 53.96 25 0Surveillance U 0.316 33.24 27.21 16 8Program Y 1.19 51.03 45.09 22 6V 2.22 59.29 46.33 25 11Weld Metal X 3.49 64.64 68.36 28 7Heat Affected U 0.316 --- 58.41 --- 13Zone Y 1.19 --- 12.98 --- 0V 2.22 --- 55.91 --- 0Material X 3.49 69.66 --- 16Notes:(a) Based on Regulatory Guide 1.99, Revision 2, methodology using the mean weightpercent values of copper and nickel of the surveillance material.(b) Calculated using measured Charpy data plotted using CVGRAPH, Version 4.1.(c) Values are based on the definition of upper shelf energy given in ASTM E185-82.(d) The fluence values presented here are the calculated values, not the best estimatevalues.Wolf Creek -Unit 116Revision 2 PRESSURE AND TEMPERATURE LIMITS REPORTTable 5.0-2Calculation of Chemistry Factors using Wolf Creek Surveillance Capsule DataMaterial Capsule Capsule FF(b) ARTNdT(c) FF*ARTNDT FF2f(a)Lower Shell U 0.316 0.684 36.46 24.94 0.468Plate R2508-3 Y 1.19 1.05 16.03 16.83 1.10(Longitudinal) V 2.22 1.22 52.03 63.48 1.49X 3.49 1.33 61.06 81.21 1.77Lower Shell U 0.316 0.684 23.79 16.27 0.468Plate R2508-3 Y 1.19 1.05 35.39 37.16 1.10(Transverse) V 2.22 1.22 54.53 66.53 1.49X 3.49 1.33 53.96 71.77 1.77SUM: 378.19 9.656CFR2508-3 = Z(FF */ARTNDT) + Y( FF2) = (378.19) + (9.656) = 39.1*FSurveillance U 0.316 0.684 18.26 (27.21) 12.49 0.468WeldMaterial(d) Y 1.19 1.05 30.26 (45.09) 31.77 1.10V 2.22 1.22 31.09 (46.33) 37.93 1.49X 3.49 1.33 45.90 (68.36) 61.05 1.77SUM: 143.24 4.828CF Smr. Weld = Z(FF

• ARTNDT) + E( FF2) = (143.24) ÷ (4.828) = 29.7*FNotes:(a) f = Calculated Fluence (1019 n/cm2, E > 1.0 MeV). These values were re-evaluated aspart of the capsule X analysis. (See Appendix D of WCAP-16028, Revision 0)(b) FF = fluence factor = fO.28-0.1Iog f)(c) ARTNDT values are the measured 30 ft-lb shift values given in the Capsule X analysisreport, WCAP-16028, [OF].(d) The Surveillance Weld ARTNDT values have been adjusted by a ratio of 0.671 (CFvw +CFsw = 32.6 ÷ 48.6). The pre-adjusted values are in parenthesis.Wolf Creek -Unit 117Revision 2 PRESSURE AND TEMPERATURE LIMITS REPORTTable 5.0-3Wolf Creek Reactor Vessel Beltline Unirradiated Material PropertiesMaterial Description Cu (%)(a) Ni(%)(a) Initial RTNDT(a)Closure Head Flange R2504-1 --- 0.66 20°F(d)Vessel Flange R2501-1 --- 0.70 20°F(d)Intermediate Shell Plate R2005-1 0.04 0.66 -20°FIntermediate Shell Plate R2005-2 0.04 0.64 -20°FIntermediate Shell Plate R2005-3 0.05 0.63 -20°FLower Shell Plate R2508-1 0.09 0.67 0°FLower Shell Plate R2508-2 0.06 0.64 10°FLower Shell Plate R2508-3 0.08(c) 0.58(c) 40°FIntermediate & Lower Plate Longitudinal 0.04(c) 0.09(c) -50OFWeld Seams (Heat # 90146)(b)Intermediate to Lower Shell Plate 0.04(c) 0.09(c) -50OFCircumferential Weld (Heat # 90146)(b)Surveillance Weld (Heat # 90146)(b) 0.06(c) 0.17(c) ---Notes:(a) Based on measured data.(b) All vessel beltline welds seams were fabricated with weld wire heat number 90146. Theintermediate to lower shell girth weld seam, 101-171, was fabricated with Flux Type 124Lot # 1061. The intermediate and lower shell longitudinal weld seams (101-124A,B, C &101-142A,B,C) were fabricated with Flux Type 0091 Lot # 0842. The surveillance weldwas fabricated with weld wire 90146, Flux Type 124 Lot # 1061. The surveillance weldmetal was made with the same weld wire heat as all the vessel beltline weld seams and istherefore representative of all the beltline weld seams.(c) Updated from previous PTS Report (WCAP-1 5080 (Ref. 8)) based on new chemicalanalysis presented in WCAP-16028.(d) These values are used for considering requirements for the heatup/cooldown curves. Perthe methodology given in WCAP-14040-A (Ref. 4), the minimum boltup temperature is600F.Wolf Creek -Unit 118Revision 2 PRESSURE AND TEMPERATURE LIMITS REPORTTABLE 5.0-4Summary of the Peak Pressure Vessel Neutron Fluence Valuesat 54 EFPY used for the Calculation of Adjusted Reference Temperature (ART) Values(n/cm2, E > 1.0 MeV)Material Surface(a) 1/4T (b) 3/4T(b)Intermediate Shell Platermediat 3.51 x 1019 2.09 x 1019 7.42x 1018R2005-1Intermediate Shell Platermediat 3.51 x 1019 2.09 x 1019 7.42x 1018R2005-2Intermediate Shell Platermediat 3.51 x 1019 2.09 x 1019 7.42x 1018R2005-3Lower Shell Plate R2508-1 3.51 x 1019 2.09 x 1019 7.42x 1018Lower Shell Plate R2508-2 3.51 x 1019 2.09 x 1019 7.42x 1018Lower Shell Plate R2508-3 3.51 x 1019 2.09 x 1019 7.42x 1018Intermediate & Lower Shell 3.08 x 1019 1.84x 1019 6.52x 101"Longitudinal Weld Seam101-124A & 101-142A (900Azimuth)Intermediate & Lower Shell 3.08 x 10'9 1.84x 1019 6.52x 1018Longitudinal Weld Seam101-124B.C &101-142B,C(2100 & 3300 Azimuth)Intermediate to Lower Shell 3.51 x 1019 2.09 x 1019 7.42x 1018Plate Circumferential WeldSeam 101 -171Notes:(a) The fluence was taken from the peak azimuthal location. (see Table 2 of WCAP-1 6030)(b) Attenuation of the fluence at the specific depth is calculated by the formula: f (depth x) = fsurace *e(°24x), where x inches (vessel beltline thickness is 8.63 inches) is the depth into the vesselwall measured from the vessel clad/base metal interface.Wolf Creek -Unit 119Revision 2 PRESSURE AND TEMPERATURE LIMITS REPORTTABLE 5.0-5Summary of Adjusted Reference Temperatures (ARTs) for the Reactor Vessel BeltlineMaterials at the '/4-T and %/4-T Locations for 54 EFPY54 EFPY ART(a)Material__ _ _ _ _ _ _ _ _ _________ _______ ___RG 1.99 Rev. 2 %-T (0F) %-T (fF)MethodIntermediate Shell Plate Position 11 42 28R2005-1Intermediate Shell Plate 42 28R2005-2Position 1-1R2005-2Intermediate Shell Plate Position 1-1 51 37R2005-3Lower Shell Plate R2508-1 Position 11 104 87Lower Shell Plate R2508-2 Position 11 88 78Lower Shell Plate R2508-3 Position 11 135 121Position 2-1 104(b) 93(b)Intermediate & Lower Plate Position 11 28 10Long. Weld Seams(Heat # 90146) Position 2-1 14 5Intermediate & Lower Plate Position 11 28 10Long. Weld Seams(Heat # 90146) Position 2-1 14 5Notes:(a) ART = Initial RTNDT + ARTNDT + Margin (OF)(b) These ART values are used to generate the heatup and cooldown curves.Wolf Creek -Unit 120Revision 2 PRESSURE AND TEMPERATURE LIMITS REPORTTABLE 5.0-6Calculation of Adjusted Reference Temperatures (ARTs) at 54 EFPY for the Limiting WolfCreek Reactor Vessel Material (Lower Shell Plate R 2508-3)Parameter ART ValueLocation 1/ 1/4-T %-TChemistry Factor, CF (OF) 39.1 39.1Fluence -1019 n/cm2 (E > 1.0 MeV), f(a) 2.09 0.742Fluence Factor, FF(b) 1.201 0.916ARTNDT = CF x FF, (OF) 46.959 35.816Initial RTNDT, I (°F) 40 40Margin, M (OF)(C) 17 17ART= I +(CFxFF)+ M (OF) 104 93per Regulatory Guide 1.99, Rev. 2Notes:(a) Fluence, f, is based upon fs,, (1019 n/cm2, E > 1.0 MeV) = 2.09 at 54 EFPY. The WolfCreek reactor vessel wall thickness is 8.625 inches at the beltline region.(b) Fluence Factor (FF) per Regulatory Guide 1.99, Revision 2, is defined asFF = fO.28 -0.1olog.f)(c) Margin is calculated as M = 2(ca2+ GA2)05.The standard deviation for the initial RTNDTmargin term ag, is 0°F since the initial RTNOT is a measured value. The standarddeviation for ARTNDT term GA, is 170F for the plate, except that GA need not exceed the0.5 times the mean value of ARTNDT.Wolf Creek -Unit 121Revision 2 PRESSURE AND TEMPERATURE LIMITS REPORTTABLE 5.0-7RTpTs Calculation for Wolf Creek Beltline Region Materials at Life Extension (54 EFPY)Material Fluence, f FF(b) CF ARTPTS Margin RTNDT(U) RTPTS(n/cm2, E>1.0 (OF) (c) (OF) (d) (OF) (e) (OF) (f) (OF) (g)MeV)(a)Intermediate Shell Plate R2005-1 3.51 x 1019 1.33 26.0 34.58 34.00 -20 49Intermediate Shell Plate R2005-2 3.51 x 1019 1.33 26.0 34.58 34.00 -20 49Intermediate Shell Plate R2005-3 3.51 x 1019 1.33 31.0 41.23 34.00 -20 55Lower Shell Plate R2508-1 3.51 x 1019 1.33 58.0 77.14 34.00 0 111Lower Shell Plate R2508-2 3.51 x 10'9 1.33 37.0 49.21 34.00 10 93Lower Shell Plate R2508-3 3.51 x 1019 1.33 51.0 67.83 34.00 40 142= Using Surv. Capsule Data 3.51 x 10"' 1.33 39.1 52.00 17.00 40 109Intermediate & Lower Plate Long. 3.51 x 1019 1.33 32.6 43.36 43.36 -50 37Weld Seams (Heat # 90146)=> Using Surv. Capsule Data 3.51 x 1019 1.33 29.7 39.50 28.00 -50 18Intermediate & Lower Plate Long. 3.51 x 1019 1.33 32.6 43.36 43.36 -50 37Weld Seams (Heat # 90146)= Using Surv. Capsule Data 3.51 x 1019 1.33 29.7 39.50 28.00 -50 18Notes:(a) The fluence was taken from the peak azimuthal location (See Table 2 of Ref. 6).(b) FF = f.28-0.1*109f); where f is the clad/base metal interface fluence.(c) Chemistry Factor is taken from Table 5 of Ref. 6.(d) ARTPTS = CF

• FF(e) Margin = 2*(au2 +CFA2)1/2.(f) Initial RTNDT values are measured values.(g) RTPTS = RTNDT(U) + ARTPTS + Margin (OF) (This value was rounded per ASTM E29, usingthe "Rounding Method")Wolf Creek -Unit 122Revision 2