ML15322A055: Difference between revisions

From kanterella
Jump to navigation Jump to search
(Created page by program invented by StriderTol)
(Created page by program invented by StriderTol)
Line 15: Line 15:


=Text=
=Text=
{{#Wiki_filter:}}
{{#Wiki_filter:Tennessee Valley Authority, Post Office Box 2000, Soddy Daisy, Tennessee 37384-2000November 13, 201510 CFR 50.4ATT-N: Document Control DeskU.S. Nuclear Regulatory CommissionWashington, D.C. 20555-0001Sequoyah Nuclear Plant, Units I and 2Renewed Facility Operating License Nos. DPR-77 and DPR-79NRC Docket Nos. 50-327 and 50-328
 
==Subject:==
Sequoyah Unit I Pressure Temperature Limits Report,Revision 5, and Sequoyah Unit 2 Pressure Temperature LimitsReport, Revision 6
 
==References:==
: 1. Letter from NRC to TVA, "Sequoyah Nuclear Plant, Units I and 2-Issuance of Amendments for the Conversion to the ImprovedTechnical Specifications with Beyond Scope Issues (TAC Nos.MF3128 and MF3129)," dated September 30, 20015(MLI15238B460)In accordance with Sequoyah Nuclear Plant (SQN) Units 1 and 2 TechnicalSpecifications (Tss) 5.6.4.c, enclosed is the Unit I Pressure Temperature LimitsReport (PTLR), Revision 5, and Unit 2 PTLR, Revision 6. In accordance with TSs5.6.4.c, the PTLRs are required to be provided to the Nuclear RegulatoryCommission (NRC) within 30 days after any revision. Sequoyah Units I and 2 wereissued license amendment Nos. 334 and 327, respectively for improved standardTSs (Reference 1). These license amendments resulted in the enclosed revisions toeach of the PTLRs. The revisions also include other editorial clarifications andadministrative changes identified during the revision process as described inEnclosure 1. The revised PTLRs became effective on October 16, 2016.There are no new regulatory commitments in this letter. If you have any questions,please contact Jonathan Johnson, SQN Site Licensing Manager at (423) 843-8129.
U.S. Nuclear Regulatory CommissionPage 2November 13, 2015Sequo Nuclear PlantEnclosures1.2.3.Units 1 and 2 Pressure Temperature Limits Report ChangesSequoyah Unit 1 Pressure Temperature Limits Report, Revision 5Sequoyah Unit 2 Pressure Temperature Limits Report, Revision 6ZTK: DVGEnclosurescc (Enclosures):NRC Regional Administrator -Region IINRC Senior Resident Inspector -SQN ENCLOSURE1ISEQUOYAH UNITS 1 AND 2PRESSURE TEMPERATURE LIMITS REPORT CHANGESThe following describes the editorial clarifications and administrative changes madeto each Units' Pressure Temperature Limits Report (PTLR).1. Conflict between the figure index referencing Cold Overpressure MitigationSystem (COMS) and the figure title referencing Low TemperatureOverpressure Protection System (LTOPS) was Corrected for consistency withthe system terminology in Section 3.4.12 of the Technical Specifications.2. Section 1, "RCS Pressure Temperature Limits Report (PTLR)," was revisedto clarify which Limiting Condition for Operations are affected by the PTLR.3. Section 3.1, "Pressurizer PORV Lift Setting Limits," was revised forconsistency with the analysis of record contained in Topical Report No.WCAP-1 5293, Revision 2 "Sequoyah Unit 1 Heatup and Cooldown LimitCurves for Normal Operation and PTLR Support Documentation," andTopical Report No. WCAP-15321 Revision 2 "Sequoyah Unit 2 Heatup andCooldown Limit Curves for Normal Operation and PTLR SupportDocumentation".4. Section 5.0, "Supplemental Data Tables," added the word "Forging" to thesentence in refer to the limiting beltline material identified in Tables 5-5 and5-6.5 A note was added to Table 4-1, "Sequoyah Unit 1 Reactor VesselSurveillance Capsule Withdrawal Schedule," to describe current and futureactivity with the surveillance capsule withdrawal schedule.6. Tables 5-5 and 5-6 were revised with a different annotation conventionconsistent with the other tables in the PTLR.
ENCLOSURE 2SEQUOYAH UNIT 1PRESSURE TEMPERATURE LIMITS REPORT, REVISION 5 B88 15 t 16800MAPRESSURE TEMPERATURE LIMITS REPORT....APPROVEDThis qp~provaIl does. not reliev, theContractor from any port o9 his re-,*ponsibility Jar, the correctness of.* design. debugailnd dimlnsio~ns.YLetter lo.Nl107l6.*oot,: ,October 16, 2015. -.SQEP BY: W. JX PierceTennessee Valley AuthoritySequoyah unit 1Pressure Temperature Limits ReportRevision 5, September 2015PROJECT Sequovah .DISCIPLINE N"CONTRACT .4411 1JNT' 1DESC. RCS Pressure-Temperature Limit Report*DWG/DOC NO. PTLR-1SHEET -OF -. "REV. 05*DATE 10/16/15 ECN/DCN -. FILE N2N-081EDMS, WT CA-K PRESSURE TEMPERATURE LIMITS REPORTTable of ContentsList of Tables .................................................................................................... ivList of Figures......................................................................................................v1.0 RCS Pressure Temperature Limits Report (PTLR) ......................................... ....... 12.0 Operating Limits.......................................................................................... 12.1 RCS Pressure/Temperature (PIT) Limits (TS 3.4.3) ............................................... 13.0 Low Temperature Overpressure Protection System (TS 3.4.12)...... ............................ 13.1 Pressurizer PORV Lift Setting Limits .............................................................. 23.2 Arming Temperature................................................................................ 24.0 Reactor Vessel Material Surveillance Program...................................................... 25.0 Supplemental Data Tables............................................................................... 36.0 References................................................................................................ 19ii PRESSURE TEMPERATURE LIMITS REPORTList of TablesTable 2-1 Sequoyah Unit 1 Heatup Limits at 32 EPPY(with Uncertainties for Instrumentation Errors of 10&deg;F and 60 psig)......... ...................... 6Table 2-2 Sequoyah Unit 1 Cooldown Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 1 0&deg;F and 60 psig)................................STable 3-1 Selected Setpoints, Sequoyah Unit 1.......................................................... 10Table 4-1 Sequoyah Unit I Reactor Vessel Surveillance Capsule Withdrawal Schedule ............ 12Table 5-1 Comparison of the Sequoyah Unit 1 Surveillance Material 30 ft-lb TransitionTemperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99,revision 2, Predictions .............................'............................................. 13Table 5-2 Calculation of Chemistry Factors using Sequoyah Unit 1 Surveillance Capsule Data. .... 14Table 5-3 Reactor Vessel Beltline Material Unirradiated Toughness Properties forSequoyah Unit 1 ............................................................................ ..... 15Table 5-4 Peak Neutron Fluence Projections at Key Azimuthal Locations on the Reactor VesselClad/Base Metal Interface for Sequoyah Unit 1 (x 10'9 n/cm2, B > 1.0 MeV) ............ 16Table 5-5 Sequoyah Unit 1 Calculation of the ART Values for the 1 /4T Location @ 32 EFPY.....17Table 5-6 Sequoyah Unit 1 Calculation of the ART Values for the 3/4T Location @ 32 EFPY ......17Table 5-7 Summary of the Limiting ART Values Used in the Generation of the Sequoyah Unit 1Heatup/Cooldo~vn Curves. .................................................................... 18Table 5-8 RTm-s Calculatiohis for Sequoyah Unit 1 Beitline Region Materials at 32 EFPY.......... 18iii PRESSURE TEMPERATURE LIMITS REPORTList of FiguresFigure 2-1 Sequoyah Unit 1 Reactor Coolant System Heatup Limitations(Heatup Rate of 1 00&deg;F/hr) Applicable for the First 32 EFPY(w/ Margins for Instrumentation Errors of 1 0&deg;F and 60 psig)................................ 4Figure 2-2 Sequoyah Unit 1 Reactor Coolant System Cooldown Limitations(Cooldown Rates up to 100&deg;F/hr) Applicable for the First 32 EFPY(w/ Margins for Instrumentation Errors of 1 0&deg;F and 60 psig)................................ 5Figure 3-1 Sequoyah Unit 1 Selected LTOPS Setpoints ................................................. 11iv PRESSURE TEMPERATURE LIMITS REPORT1.0 RCS Pressure Temperature Limits Report (PTLR)This PTLR for Sequoyah Unit 1 has been prepared in accordance with the requirements of TecirnicalSpecification (TS) 5.6.4. Revisions to the PTLR shall be provided to the NRC after issuance.This report affects TS 3.4.3, RCS Pressure/Temperature Limits (PiT) Limits and TS 3.4.12, LowTemperature Over Pressure Protection (LTOP) System.2.0 RCS Pressure and Temperature LimitsThe limits for TS 3.4.3 are presented in the subsections which follow and were developed using the NRCapproved methodologies specified in TS 5.6.4 with exception of ASME Code Case N-640[13] (Use of Kit),WCAP- 15984-P['14 (Elimination of the Flange Requirement), 1996 Version of Appendix Gm and the-revised fluencesE7T. The operability requirements associated with LTOPS are specified in TS 3.4.12 andwere determined to adequately protect the RCS against brittle fracture in the event of an LTOP Transientin accordance with the methodology specified in TS 5.6.4.2.1 RCS Pressure/Temperature (PIT) Limits (TS 3.4.3)2.1.1 The minimum boltup temperature is 500F2.1.2 The RCS temperature rate-of-change limits are:a. A maximum heatup rate of 1 00&deg;F in any one hour period.* b. A maximum cooldown rate of l00OF in any one hour period.c. A maximum temperature change of less than or equal to I10&deg;F in any one hour period duringinservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.2.1.3 The RCS P/T limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticalityare specified by Figures 2-1 and 2-2.3.0 Low Temperature Overpressure Protection System (TS 3.4.12)The lift setpoints for the pressurizer Power Operated Relief Valves (P.ORVs) are presented in thesubsections which follow. These lift setpoinlts have been developed using the NRC-approvedmethodologies specified in TS 5.6.4.1 PRESSURE TEMPERATURE LIMITS REPORT3.1 Pressurizer PORV Lift Setting LimitsThe pressurizer PORV lift setpoints are specified by Figure 3-1 and Table 3-1 (Ref. 10). The limits for theLTOPS setpoints are contained in the 32 EFPY steady-state curves (Table 2-2), which are beitlineconditions and are not compensated for pressure differences between the pressurizer transmitter and thereactor midplane/beltline or for instrument inaccuracies. The pressure difference between the pressurizertransmitter and the reactor vessel midplane/beltline with four reactor coolant pumps in operation is 68.3psi (Ref. 11).Note: These setpoints include allowance for the pressure difference between the pressurizertransmitter and the reactor vessel midplane/beltline and the SOTF thermal transport effect forheat injection transients. A demonstrated accuracy calculation (Reference 12) has beenperformed to confirn that the setpoints will maintain the system pressure within theestablished limits when the pressure difference between the pressure transmitter and reactormidplane and maximum temperature/pressure instrument uncertainties are applied to thesetpoints.3.2 Anning TemperatureThe LTOPS arming temperature is based upon the methodology defined in the Sequoyah Nuclear PlantUnit 1 Technical Specifications Administrative Controls Section 5.6.4. The arming temperature shall be _<350&deg;F.4.0 Reactor Vessel Material Surveillance ProgramThe reactor vessel material irradiation surveillance specimens shall be removed and examined todetermine changes in material properties. The removal schedule is provided in Table 4-1. The results ofthese examinations shall be used to update Figures 2-1, 2-2 and 3-1.The pressure vessel steel surveillance program (WCAP-8233 r1l) is in compliance with Appendix H to 10CFR 50, "Reactor Vessel Material Surveillance Program Requirementsr21."' The material test requirementsand the acceptance standard utilize the reference nil-ductility temperature which is determined inaccordance with ASTM E23 [3]. The empirical relationship between RTNDr and the fracture toughness ofthe reactor vessel steel is developed in accordance with Code Case N-640 of Section XI of the ASMEBoiler and Pressure Vessel Code, Appendix 0, "Fracture Toughness Criteria for Protection AgainstFailureE41.'' The surveillance capsule removal schedule meets the requirements of ASTM E185-82N5.Theremoval schedule is provided in Table 4-1.2 PRESSURE TEMPERATURE LIMITS REPORT5.0 Supplemental Data TablesTable 5-1 contains a comparison of measured surveillance material 30 ft-lb transition temperature shiftsand upper shelf energy decreases with Regulatory Guide 1.99, Revision 2[6], predictions.Table 5-2 shows calculations of the surveillance material chemistry factors using surveillance capsuledata. Note that in the calcuilation of the surveillance weld chemistry factor, the ratio procedure fromRegulatory Guide 1.99, Revision 2 was followed. The ratio in question is equal to 0.90.Table 5-3 provides the required Sequoyah Unit 1 reactor vessel toughness data.Table 5-4 provides a summary of the fluence values used in the generation of the heatup and cooldownlimit curves and the PTS evaluation.Table 5-5 and 5-6 show the calculation of the 1/4T and 3/4T adjusted reference temperature at 32 EFPYfor each beitline material in the Sequoyah Unit 1 reactor vessel. The limiting beitline material was theLower Shell Forging 04.Table 5-7 provides a summary of the adjusted reference temperature (ART) Values of the Sequoyah Unit 1reactor vessel beltline materials at the 1/4T and 3/4T locations for 32 EFPY.Table 5-8Sprovides RTPTs values for Sequoyah Unit 1 at 32 EFPY.3 PRESSURE TEMPERATURE LIMITS REPORTMATERIAL PROPERTY BASISLTMITING MATERIAL: LOWER SHELL FORGING 04LIMITING ART VALUES AT 32 EFPY: 1/4T, 216&deg;F3/4T, 186&deg;F2500 --_o- erlim Version.5.1 Run.15680 [i /-2250i Leak Test Limit /-; -2000 .......Unacceptable ....... ... -- --"Acceptable ...,Operation, Operation__:Heatup Rate Critical Limit100o Deg. F/HrI 100 Deg. F/ HrI-- "~2=i(n 'i.120 ,.. .. ....S 1000 --- _ __- ___ __ _750 -- -- -__ _ _ -__ ___-__ __ __ __ --- __/ Criticality Limit based on Iinservice hydrostatic test500 -temnperatuire (2880F) for the __service period up to 32 EFPY(IMinimum !250 ......-~ Boltup .. ... -; -___ __) _STemp = 50*F0 50 100 150 200 250 300 350 400 450 500 550Moderator Temperature (Deg. F)Figure 2-1 Sequoyah Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of100&deg;F/hr)-Applicable for the First 32 EFPY (w/Margins for Instrumentation Errorof 10&deg;F and 60 psig) (Plotted Data provided on Table 2-1)4 PRESSURE TEMPERATURE LIMITS REPORTMATERIAL PROPERTY BASISLIMITING MATERIAL: LOWER SHELL FORGING 04LIMITING ART VALUES AT 32 EFPY: 114T, 216&deg;F3/4T, 186&deg;F2500225020001750150021250o 10007505002500 50 100 150 200 250300 350 400 450 500 550Moderator Temperature (Deg. F)Figure 2-2 Sequoyah Unit 1 Reactor Coolant System Cooldown Limitations (CooldownRates up to 100&deg;F/hr) Applicable for the First 32 EFPY (w/Margins forInstrumentation Error of 10&deg;F and 60 psig) (Plotted Data provided on Table 2-2)5 PRESSURE TEMPERATURE LIMITS REPORTTable 2-1Sequoyah Unit 1 Heatup Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 10&deg;F and 60 psig)100 Heatup 1100 Critica] Limit ILeak Test LimitT P J T Pj T P50 0 288 0 272 200050 477 *288 477 288 248555 477 288 47760 477 288 47765 477 288 47770 477 288 47875 477 288 47880 477 288 48085 477 288 48190 477 ,288 48395 477. 288 485100 477 288 487105 477 288 490110 477 288 493115 477 288 497120 477 288 500125 477 288 505"130 477 288 508135 477 288 515140 477 288 517145 .477 288 527150 477 288 528155 478 288 541160 480 288 541165 483 288 555170 487 288 557175 493 288 571180 500 288 575185 508 288 589190 517 288 609195 528 288 631200 541 288 .656205 555 ,288 684210 571 288 714215 589 .288 748220 609 290 786225 631 295 8286 PRESSURE TEMPERATURE LIMITS REPORTTable 2-1 -(Continued)Sequoyah Unit 1 Heatup Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 1 0&deg;F and 60 psig)100 Heatup 100 Critical LimitT P T P230 656 300 874235 684 305 925240 714 310 981245 748 315 1044250 786 320 1112255 828 325 1188260 874 330 1272265 925 335 1364270 981 340 1466275 1044 345 1578280 1112 350 1702285 1188 355 1838290 1272 360 1988295 1364 365 2154300 1466 370 2337305 1578310 1702315 1838320 1988325 2154330 23377 PRESSURE TEMPERATURE LIMITS REPORTTable 2-2Sequoyah Unit 1 Cooldown Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 10&deg;F and 60 psig)Steady State [20F )40F P1 60F 100F50505560657075808590951001051101151201251301351401451501551601651701751801851901952002052102150552553555556558560561564566569571575578582586591596602608616623632642652664677691707724743764788814843505055606570758085,909510010511011512012513013514014515015516016517017518018519019520020521021505035055075095105125145165185215245275315355405455505565635715795885996106236376526696887097337597878195050556065707580859095100105110115120125130135140145150155160165170175180185190195200205210215045745845946046246446546847047347647948348749249750351051752553454455656858259761463365467770273176279750505560657075808590951001051101151201251301351401451501551601651701751801851901952002052102150408409410411412414416418420423426430434.4384434494564634714794895005125265415585775976206466747057407795050556065707580859095100105110115120125130'1351401451501551601651701751801851901952002052102150305306307308309311.313315318.321325329333338344351358367376"387399412427443461482505530558590624663706754______________________________________ +/- _______________________________________ _________________________________________ I8 PRESSURE TEMPERATURE LIMITS REPORTTable 2-2 -(Continued)Sequoy~a Unit 1 Cooldown Limits at 32 EFPY(without Uncertainties for Instrumentation Errors)Steady State 20F 40F 60F 100F.T P T P T PT P T P220 874 220 853 220 836 220 821 220 806225 909 225 892 225 878 225 869 2-25 865230 948 230 935 230 925 230 921235 991 235 982 235 978240 1038 240 1034245 1090250 1148255 1212260 1283265 1360270 1447275 1542280 1647285 1763290 1892295 2034300 2191305 23649 PRESSURE TEMPERATURE LIMITS REPORT*Table 3-1Selected Setpoints, Sequoyah Unit 1Trs Dg.) PORV#2 PORV#1Trcs(De.F)Setpoint (psig) Setpoint (psig)50 490 465100 500 475135 540 510175 575 540200 610 570250 745 685280 745 685405 745 685450 2350 235010 PRESSURE TEMPERATURE LIMITS REPORTSequoyah Unit I LTOPS Selected Setpoints250020002i1500100005000 50 100 150 200 250 300 350 400 450 500Reactor Coolant System Temnperature (&deg;F)--4-- FOFRV#2 Setpoint --=- PORV#1 SetpointFigure 3-1 Sequoyah Unit 1 Selected LTOPS Setpoints (Plotted Data provided on Table 3-1)11 PRESSURE TEMPERATURE LIMITS REPORTTable 4-1Seqtuoyah Unit 1 Reactor Vessel Surveillance Capsule Withdrawal Schedule* Removal Time FluenceCapsule Location Lead Factorca) (EFPY) (b) (n/cm2,E>l.0 1T 400 3.39 1.03 2.61 x 10O8 (c)U 1400 3.47 3.00 7.96 x 1018 (c)X 2200 3.47 5.27 1.32 x 1019 (c)Y 320&deg; 3.43 10.03 2.19 x 10'9 (c,d)5 40 1.08 Standby (d,e)V 176&deg; 1.08 Standby (d,e)W 184&deg; 1.08 Standby (d,e)Z 3560 1.08 Standby (d,e)Notes:(a)(b)(c).(d)(e)Updated in Capsule Y dosimetry analysis Effective Full Power Years (EFPY) from plant startup.Plant specific evaluation.This fluence is not less than once or greater than twice the peak end of license (32 EFPY) fluenceCapsules 5, V, W and Z will reach a fluence of 2.74 x 1019 (E > 1.0 Mev), the 48 EFPY peakvessel fluence at approximately 44 EFPY, respectively.Administrative Note -The surveillance capsule withdrawal schedule in Table 4-1 is based on thesurveillance program for the original 40 year service life. Relocation of select standby capsules toincrease the fluence lead factor in anticipation of the updated surveillance program for the 60 year licenserenewal service life is described in a TVA letter to NRC dated May 14, 2015 (ML15!34A377).Regulatory approval for the anticipatorY standby capsule relocation has been granted by a NRC letter toTVA dated September 4, 2015 (ML 15244B222). A complete update of the reactor vessel surveillanceprogram for the 60 year license renewal service life will be documented by a subsequent revision to thePTLR prior to entry into the license renewal extended operating period.12 PRESSURE TEMPERATURE LIMITS REPORTTable 5-1Comparison of the Sequoyah Unit 1 Surveillance Material 3 0 ft-lb Transition Temperature Shifts andUpper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions30 ft-lb Transition Upper Shelf EnergyTemperature Shift DecreaseMaterial Capsule Fluence Predicted Measured Predicted Measured____________ ______ (X 1019 n/cm2). (OF)(a) (OF)(b) (%)(a) .(%)(C)Lower Shell T 0.261 .59.85 67.52 16 16Forging 04 U 0.796 89.3 109.7 20.5 21(Tneta)X 1.32 102.6 145.12 23 8(Heat # 980919 /281587) Y 2.19 114.95 129.87 26.5 23Lower Shell T 0.261 59.85 50.59 16 0Forging 04 U 0.796 89.3 67.59 20.5 19(xa)X 1.32 102.6 103.34 23 22(Heat # 980919 / _____281587) Y 2.19 114.95 133.35 26.5 19Weld Metal T 0.261 111.13 127.79 .35 30(Heat # 25295)(d) U 0.796 165.82 144.92 42 26X 1.32 190.51 159.02 .45 .21Y 2.19 213.44 163.8 48 28IiAZ Metal T 0.261 --45.48 --20U 0.796 --. 78.94 --26X 1.32 --95.89 --3Y 2.19 --73.3 --10Notes:(a) Based on Regulatory Guide 1.99, Revision 2, methodology using the mean weight percent values ofcopper and nickel of the surveillance material.(b) .Calculated using measured Charpy data plotted using CVGRAPH, Version 4.1 [8(c) Values are based on the definition of upper shelf energy given in ASTM E185-82.(d) Surveillance Weld was fabricated from weld wire type SMIT 40, Heat # 25295, Flux type SMIT 89,Lot # 1103.13 PRESSURE TEMPERATURE LIMITS REPORTTable 5-2Calculation of Chemistry Factors using Sequoyah Unit 1 Surveillance Capsule DataMaterial Capsule Capsule f~a) FF) ARTNBT(C) FF*ARTNDT FF2Lower Shell T 2.61E+18 0.63 67.52&deg;F 42.54&deg;F 0.40Forging 04 U 7.96E+18 0.94 109.7&deg;F 103.12&deg;F 0.88(Tangential)X 1.32E+19 1.08. 145.12&deg;F 156.73&deg;F 1.16(Heat # 980919/ /_________281587) Y 2.19E+19 1.21 129.87&deg;F 157.14&deg;F 1.47Lower Shell T 2.6 1E+I18 0.63 50.59&deg;F 31.87&deg;F 0.40Forging 04 U 7.96E+18 0.94 67.59&deg;F 63.53&deg;F 0.88(Axial) X 1.32E+19 1.08 103.34&deg;F 111.61&deg;F 1.16(Heat #980919 / *Y 2.19E+19 1.21 133.35&deg;F 161.35&deg;F 1.47281587)SUM: 827.89&deg;F 7.82CF04 =
* RTNDTr) + 2( FF2) = (827.89) +(7.82) = 105.9&deg;FSurveillance Weld T 2.61E+18 0.63 115.0&deg;F 72.5&deg;F 0.40Material(d) U 7,96E+18 0.94 130.4&deg;F 122.6&deg;F 0.88(Heat # 25295)(e) X 1.32E+19 1.08 143.1&deg;F 154.5&deg;F 1.16Y 2.19E+19 1.21 147.4OF 178.40F 1.47SUM: 528.0&deg;F 3.91CF Surv. we~d =XY(FF
* RTrrDT) +X( FF2) =(528.0&deg;F) + (3.91) = 135.0&deg;FNotes:(a) f= Calculated fluence from Capsule Y dosimetry analysis resultsETI, (n/cm2, E > 1.0 MeV).(b) FF = fluence factor =fI0.28"0.1logf.(c) ARTNDTValues are the measured 30 ft-lb shift values taken from App. B of Ref. 7, rounded to onedecimal point.(d) The surveillance weld metal ARTNDoT values have been adjusted by a ratio factor of 0.90.(e) Surveillance Weld was fabricated from weld wire type SMIT 40, Heat # 25295, Flux type SMIT 89,Lotft 110314 PRESSURE TEMPERATURE LIMITS REPORTTable 5-3Reactor Vessel Beltline Material Unirradiated Toughness Properties for Sequoyah Unit 1Material Description Cu (%) Ni (%) Initial RTNDT(a)Intermediate Shell Forging 05(Heat'#980807/281489) 01 .640Lower Shell Forging 04(Heat #980919/281587) 01 .670Surveillance Weld (Heat # 25295)(b'd, e) == 0.387 0.11 ---Rotterdam Test(c. e) 0.30 ......-Rotterdam Test(c' e) 0.25 ......Rotterdam Test(c' e) 0.46 ......-Best Estimate of the Intermediate to Lower ShellForging Circumferential Weld Seam W05 0.35 0.11 -40OF(Heat # 25295)(d. e)(a) The Initial RTNDT values are measured values(b) These copper and nickel values are best estimate values for only the surveillance weld metal and is the averageof three data points [0.424 (WCAP-10340, Rev.1), 0.406 (WCAP-10340, Rev.1), 0.33 (WCAP-8233) copperand 0.084 (WCAP-10340, Rev.1), 0.085 (WCAP-10340, Rev.1), 0.17 (WCAP-8233) nickel.]. These values aretreated as one data point in the calculation of the best estimate average for the inter, to lower shell circ. weldshown above. Originally the 0.424 / 0.406 and 0.084 / 0.085 values were reported as single points, 0.41 -0.42and 0.08 (Per WCAP-10340, Rev. l[7d]), but it is actually made up of two data points. Sample TW58 fromCapsule T was broken into two samples, TW58a and TW58b, thus providing the two data points.(c) From NRC Reactor Vessel Integrity Database (RVID) and ultimately fr'om Rotterdam Weld Certifications.(d) Circumferential Weld Seam W05 was fabricated with weld wire type SMvIT 40, Heat # 25295, Flux type SMVIT89, Lot # 2275. The surveillance weld was fabricated with weld Wire type SMIT 40, Heat # 25295, Flux typeSMIT 89, Lot # 1103 and is representative of the intermediate to lower shell circumferential weld.(e) The surveillance weld and the three Rotterdam tests are averaged together for the Best Estimate of theIntermediate to Lower Shell Forging Circumferential Weld Seam.15 PRESSURE TEMPERATURE LIMITS REPORTTable 5-4Peak Neutron Fluence Projections at Key Azimuthal Locations on the Reactor Vessel Clad/Base MetalInterface for Sequoyah Unit 1 (x 10 9 n/cm2, E > 1.0 MeV)Azimuthal LocationEFPY 00 150 300 45010.03 0.205 0.321 0.409 0.63720 0.387 0.596 0.761 1.1832 0.605 0.928 1.19 1.8448 0.896 1.37 1.75 2.7216 PRESSURE TEMPERATURE LIMITS REPORTTable 5-5Sequoyah Unit 1 Calculation of the ART Values for the 1/4T Location @ 32 EFPY(a)Material RG 1:99 CF FF ARTnDT(C) Margin(d) ART(e)R2 Method (0F) (&deg;F) (0F) (0F) (0F)Intermediate Shell Forging 05 Position 1.1 115.6 1.029 40 119.0 34 193Position 1.1 95 1.029 73 97.8 34 205Lower Shell Forging 04Position 2.1 105.9 1.029 73 109.0 34(0 216Intermediate to Lower Shell Position 1.1 161.3 1.029 -40 166.0 56 182Circumferential Weld Seam Position 2.1 135.0 1.029 -40 138.9 56(0 155Notes:(a) Neutron fluence value used for all materials is the highest value from Table 5-4 for 32 EFPY.(b) Initial RTNIJT values are measured values.(c) AXRTNDT = CF
* FF(d) M = 2 *(a'2 +- OA2)1/2(e) ART = Initial RTNDT + ARTNDT + Margin (0F)(f) Data deemed not-credible (See Reference 7a), thus the full GA will be used to determine margin.Table 5-6Sequoyah Unit 1 Calculation of the ART Values for the 3/4T Location @ 32 EFPY(a)Material RG 1.99 CF FF IRTrmT(b) ART~rTC() Margin(d) ART(e)*R2 Method (0F) (0F) (0F) (0F) (0F)Intermediate Shell Forging 05 Position 1.1 115.6 0.747 40 86.4 34 160Loe hl ogn 4 Position 1.1 *95 0.747' 73 71.0 34 178*Position 2.1 105.9 0.747 73 79.1
* 34(0 186Intermediate to Lower Shell Position 1.1 161.3 0.747 -40 120.5 56 137Circumferential Weld Seam Position 2.1
* 135.0 0.747 -40 100.8 56(0 117Notes:(a) Neutron fluence value used for all materials is the highest value from Table 5-4 for 32 EFPY.(b) Initial RTNDT values are measured values.(c) ARTrDT =CF *FF(d) M = 2 *(a.2 + Gra2)1t2(e) ART = Initial RTNDT .+- ARTrNDT + Margin (0F)(f) Data deemed not-credible (See Reference 7a), thus the full GrA will be used to determine margin.17 PRESSURE TEMPERATURE LIMITS REPORTTable 5-7Summnary of the Sequoyah Unit 1 Reactor Vessel Beitline Material ART ValuesMaterial RG 1.99 R2 1/4 ART 3/4 ARTMethod (0F) (0F)Intermediate Shell Forging 05 Position 1.1 193 160Position 1.1 205 178Lower Shell Forging 04Position 2.1 216 186Intermediate to Lower Shell Position 1.1 182 137Circumferential Weld Seam Position 2.1 155 117Table 5-8RTP-s Calculations for Sequoyah Unit 1 Beltline .Region Materials at 32 EFPY(a)Material Fluence FF CF ARTpTs~b) Margin RTpxs(d)(X 1019 n/cm2, "(F) (0F) (0F) (OF) (0F)E>1.0 MeV)Intermediate Shell Forging 05 1.84 1.167 115.6 .134.9 34 40 209Lower Shell Forging 04 1.84 1.167 95.0 110.9 34 73 218Lower Shell Forging 04 1:84 1.167 105.9 123.6 34(e) 73 231(Using S/C Data)Circumferential Weld Metal 1.84 1.167 161.3 188:2 56 -40 204Circumferential Weld Metal 1.84 1.167 135.0 157.5 56(e) -40 174(Using S/C Data)Notes:(a) Neutron fluence value used for all materials is the highest value from Table 5-4 for 32 EFPY.(b) ARTPTs = CF *FF(c) Initial RTNOT values are measured values(d) RTPTs= + ARTpTs + Margin (0F)(e) Data deemed not-credible (See Reference 7a), thus the full will be used to determine margin.18 PRESSURE TEMPERATURE LIMITS REPORT6.0 References1. WCAP-8233, Tennessee Valley Authority Sequoyah Unit No. 1 Reactor Vessel RadiationSurveillance Progr~am, S. E. Yanichko, et. al., December 1973.2. Code of Federal Regulations, 10OCFR50, Appendix H, Reactor Vessel Material SurveillancePro gramn Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C..3. ASTM E23 Standard Test Method Notched Bar himpact Testing of Metallic Materials, in ASTMStandards, American Society for Testing and Materials, Philadelphia, PA.4. Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G, Fracture Toughness"Criteria for Protection Against Failure5. ASTM E185-82, Annual Book of ASTM Standards, Section 12, Volume 12.02, Standard Practicefor Conducting Surveillance Tests for Light- Water Cooled Nuclear Power Reactor Vessels.6. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S.Nuclear Regulatory Commission, May 1988.7a. WCAP-15224, Analysis of Capsule the Tennessee Valley Author"ity Sequoyah Unit 1Reactor Vessel Radiation Surveillance Program, T.J. Laubham, et. al., June 1999.7b. WCAP-1 3333, Analysis of Capsule Xfoino the Tennessee Valley Authority Sequoyah Unit 1Reactor Vessel Radiation Surveillance Progr~am, M.A. Ramirez, S. L. Anderson, L. Albertin, June1992.7c. .SwRI Project 06-8851, Reactor Vessel Material Surveillance Progr'am for Sequoyah Unit No. 1:Analysis of Capsule U, P. K. Nair, et al., October 1986.7d. WCAP- 10340, Revision 1, Analysis of Capsule T foin the Tennessee Valley Authority SequoyahUnit.] Reactor Vessel Radiation Surveillance Program, S.E. Yanichko, et. al., February 1984.8. CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.1, developed by ATIConsulting, March 1999.9. WCAP-14040-NP-A, Revision 2, Methodology Used to Develop Cold Overpressure MitigatingSystem Setpoints and RCS Heatup and Cooldown Limit Curves, J.D. Andrachek, et. al., January1996.10. WCAP-15293, Revision 2, Sequoyah Unit 1 Heatup and Cooldown Limit Curves for NormalOperation and PTLR Support Documentation, J.H. Ledger, July 2003.11. Westinghouse Letter to TVA, TVA-93 -105, Cold Overpressure Mitigation System Code Case andDelta-P Calculation, dated May 19, 1993.12. Calculation SQN-IC-01 4, Demonstrated Accuracy Calculation for Cold Overpressure ProtectionSystem.19 PRESSURE TEMPERATURE LIMITS REPORT13. ASME Code Case N-640, Alternative Reference Fracture Toughness for Development of P-TLimit Curves for Section XL, Division 1, dated February 26, 1999.14. WCAP-15984-P, Revision 01, Reactor Vessel Closure Head/Vessel Flange RequirementsEvaluation for Sequoyah Units 1 and 2, W. Bamford, et.al., April 2003.20 ENCLOSURE 3SEQUOYAH UNIT 2PRESSURE TEMPERATURE LIMITS REPORT, REVISION 6 Bi88 151.016"PRESSURE TEMPERATURE LIMITs REPORT801IContractrfo ny port of his re-.dligdealsad dinmsnsions. .*: 'Octobe~r 16,.2015. _* SOEP (Nt) BYW. J.Pierce__*Tennessee Valley AuthoritySequoyah unit 2Pressure Temperature Limits ReportRevision 6, september 2015PROJECT Secjuovah 'DisCIPLINE NCONTRACT " 4411 UNIT. 2DESC. RCS Pressure-Temperature L'mait-ReportDWG/DOC NO. PTLR-2SHEET ..- OF -REV. 06*DATE. 10/16/15 ECN/DCN FILE N2N-081"EDMS, WT CA-K PRESSURE TEMPERATURE LIMITS REPORTTable of ContentsLito T b e .List. .................of.............Tables........... .......................iList of Figures ............................................................................................. ........ v1.0 RCS Pressure Temperature Limits RePort (PTLR) ........ .... .................................... 12.0 Operating Limits ............................................................................. .............12.1 RCS Pressure/Temperature (PIT) Limits (TS 3.4.3)................................................. 13.0 Low Temperature Overpressure Protection System (TS 3.4.12) ...................... ..13.1 Pressurizer PORV Lift Setting Limits ........... .....................3.2. Arming Temperature ............'......24.0 Reactor Vessel Material Surveillance Program.....,................................................. 25.0 Supplemental Data Tables ............................................................................3..36.0 References ......................................... .. ... ..... ........ 17ii PRESSURE TEMPERATURE LIMITS REPORTList of TablesTable 2-1 Sequoyah Unit 2 Heatup Limits at 32 BEFPY(with Uncertainties for Instrumentation Errors of l0&deg;F and 60 psig)............................... 6Table 2-2 Sequoyah Unit 2 Cooldown Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 10&deg;F and 60 psig)...... ......................... 7Table 3.-1 Selected Setpoints, Sequoyah Unit 2........................................................... 8Table 4-1 Sequoyah Unit 2 Reactor Vessel Surveillance Capsule Withdrawal Schedule............. 10Table 5-1 Comparison of the Sequoyah Unit 2 Surveillance Material 30 ft-lb TransitionTemperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99,revision 2, Predictions................ .......................................................... 11Table 5-2 Calculation of Chemistry Factors using Sequoyah Unit 2 Surveillance Capsule Data....12Table 5-3 Reactor Vessel Beltline Material Unirradiated Toughness Properties forSequoyah Unit2 ................................................................................ 13Table 5-4 Peak Neutron Pluence Projections at Key Azimuthal Locations on the Reactor VesselClad/Base Metal Interface for Sequoyah Unit 2 (x 1019 n/cm2, EB> 1.0 MeV) ....i........ 14Table 5-5 Sequoyah Unit 2 Calculation of the ART Values for the 1/4T Location @ 32 EFPY.....15Table 5-6 Sequoyah Unit 2 Calculation of the ART Values for the 3/4T Location @ 32 EPPY.....15Table 5-7 Summary of the Limiting ART Values Used in the Generation of the Sequoyah Unit 2Heatup/Cooldown Curves.......................................................... :............ 16Table 5-8 RTPTs Calculations for Sequoyah Unit 2 Beltline Region Materials at 32 EFPY.......... 16iii PRESSURE TEMPERATURE LIMITS REPORTList of FiguresFigure 2-1 Sequoyah Unit 2 Reactor Coolant System Heatup Limitations(Heatup Rate of I 00&deg;F/hr) Applicable for the First 32 EFPY(wi/Margins for'Instrumentation Errors of 1 00F and 60 psig)................................ 4Figure 2-2 Sequoyah Unit 2 Reactor Coolant System Cooldown Limitations(Cooldown Rates up to 1 00&deg;F/hr) Applicable for the First 32 EFPY(w/ Margins for Instrumentation Errors of 10&deg;F and 60 psig) .................... .............5Figure 3-1 Sequoyah Unit 2 Selected LTOPS Setpoints .................................................. 9iv PRESSURE TEMPERATURE LIMITS REPORT1.0 RCS Pressure Temperature Limits Report (PTLR)This PTLR for Sequoyah Unit 2 has been prepared in accordance with the requirements of TechnicalSpecification (TS) 5.6.4. Revisions to the PTLR shall be provided to the NRC after issuance.This report affects TS.3.4.3, RCS Pressure/Temperature Limits (P/T) Limits and TS 3.4.12,.LowTemperature Overpressure Protection (LTOP) System.2.0 RCS Pressure and Temperature LimitsThe limits for TS 3.4.3 are presented in the subsections which follow and were developed using the NRCapproved methodologies specified in TS 5.6.4 with exception of ASME Code Case N-64011n1 (Use of K1o),WCAP- 159 84-Ptl21 (Elimination of the Flange Requirement), 1996 Version of Appendix G[41 and therevised fluencesETI. The operability requirements associated with LTOPS are specified in TS 3.4.12 andwere determined to adequately protect the RCS against brittle fracture in the event of an LTOP Transientin accordance with the methodology specified in TS 5.6.4.2.1 RCS Pressure/Temperature (P/T) Limits (TS 3.4.3)2.1.1 The minimum boltup temperature is 50&deg;F2.1.2 The RCS temperature rate-of-change limits are:a. A maximum heatup rate of 1 00&deg;F in any one hour period.b. A maximum cooldown rate of 100&deg;F in any one hour period.c. A maximum temperature change of less than or equal to. 1 0&deg;F in any one hour period duringinservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.2.1.3 The RCS P/T limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticalityare specified by Figures 2-1 and 2-2.3.0 Low Temperature Overpressure Protection System (TS 3.4.12)The lift setpoints for the pressurizer Power Operated Relief Valves (PORVs) are presented in thesubsections which follow. These lift setpoints have been developed using the NRC-approvedmethodologies specified in TS 5.6.4.I PRESSURE TEMPERATURE LIMITS REPORT*3.1 Pressurizer PORV Lift Setting LimitsThe pressurizer PORV lift setpoints are specified by Figure 3-1 and Table 3-1 (Ref. 10). The limits for theLTOPS setpoints are contained in the 32 EFPY steady-state curves (Table 2-2), which are beitlineconditions and are not compensated for pressure differences between the pressurizer transmitter and thereactor midplane/beltline or for instrument inaccuracies. The pressure difference between the pressurizertransmitter and the reactor vessel midplane/beltline with four reactor coolant pumps in operation is 68.3psi (Ref. 13).Note: These set-points include allowance for the pressure difference between the pressurizertransmitter and the reactor vessel midplane/beltline and the 50&deg;F thermal transport effect forheat injection transients.. A demonstrated accuracy calculation (Reference 14) has beenperformed to confirm that the setpoints will maintain the system pressure within theestablished limits when the pressure difference between the pressure transmhitter and reactormidplane and maximum temperature/pressure instrument uncertainties are applied to thesetpoints.3.2 Arming TemperatureThe LTOPS arning temperature is based upon the methodology defined in the Sequoyah Nuclear PlantUnit 2 Technical Specifications Administrative Controls Section 5.6.4. The arming temperature shall be <350&deg;F.4.0 Reactor Vessel Material Surveillance ProgramThe reactor vessel material irradiation surveillance specimens shall be removed and examined todetermine changes in material properties. The removal schedule is provided in Table 4-1. The results ofthese examinations shall be used to update Figures 2-1, 2-2 and 3-1.The pressure vessel steel surveillance program (WCAP-85 13[1]) is in compliance with Appendix H to 10CFR 50, "Reactor Vessel Material Surveillance Program Requirementst21."' The material test requirementsand the acceptance standard utilize the reference nil-ductility temperature RTNDT, which is determined inaccordance with ASTM E23 [3. The empirical relationship between RTNOT and the fracture toughness ofthe reactor vessel steel is developed in accordance with Code Case N-640 of Section XI of the ASMEBoiler and Pressure Vessel Code, Appendix G, "Fracture Toughness Criteria for Protection AgainstFailurer4k.' The surveillance capsule removal schedule meets the requirements of ASTM E185-82N5.Theremoval schedule is provided in Table 4-1.2 PRESSURE TEMPERATURE LIMITS REPORT5.0 Supplemental Data TablesTable 5-1 contains a comparison of measured surveillance material 30 ft-lb transition temperature shiftsand upper shelf energy decreases with Regulatory Guide 1.99, Revision 2[6], predictions.Table 5-2 shaows calculations of the surveillance material chemistry factors using surveillance capsuledata. Note that in the calculation of the surveillance weld chemistry factor, the ratio procedure fromRegulatory Guide 1.99, Revision 2 was followed. The ratio in question is equal to 0.93.Table 5-3 provides the required Sequoyah Unit 2 reactor vessel toughness data.Table 5-4 provides a summary of the fluence values used in the generation of the heatup and cooldownlimit curves and the PTS evaluation.Table 5-5 and 5-6 show the calculation of the 1/4T and 3/4T adjusted reference temperature at 32 EFPYfor each beltline material in the Sequoyah Unit 2 reactor vessel. The limiting beltline material was theIntermediate Shell Forging 05..Table 5-7 provides a summary of the adjusted reference temperature (ART) values of the Sequoyah Unit 2reactor vessel beltline materials at the l/4T and 3/4T locations for 32 EFPY.Table 5-8 provides RTP-Ts values for Sequoyah Unit 2 at 32 EFPY.3 PRESSURE .TEMPERATURE LIMITS REPORTMATERIAL PROPERTY BASISLIMITING MATERIAL! INTERMEDIATE SHELL FORGING 05LIMITING ART VALUES AT 32 EFPY: 1/4T, 142&deg;F3/4T, l15&deg;F2Rflfl-v Io)perlim Version:5.1 Run:5694 /ILeak, Test, Limit 22502000--[ Unacceptable1Oeain ....*Acceptable ...Operation1750 -..1500125U) 00Co0 25~ritical Limit*100 Deg. F/Hr750"500250MinumumBoltup Temp-- = 500F "Criticality Limit based on-~inservic e hydrostatic testtemperature (214&deg;F) for theservice period up to 32 EFPYUf.......I ..............! ....! ........ .........0 50 100 150 200 250 300 350 400 450 500 550Moderator Temperature (Deg. F)Figure 2-1 Sequoyah Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of1000F/hr) Applicable for the First 32 EFFY (w/ Margins for InstrumentationError of 10&deg;F and 60 psig) (Plotted Data provided on Table 2-1)
PRESSURE TEMPERATURE LIMITS REPORTMATERIAL PROPERTY BASISLIMITING MATERIAL: INTERMEDIATE SHELL FORGING 05.LIMITING ART VALUES AT 32 EFPY: 1/4T, 142&deg;F3/4T, 150F.2500 [Oeri Version:5.1 Run:5694]2000 tUnacceptable _____ --__ _________ AcceptableOperation OPeration -1750 ..... 'S 1250 * -.-o----__ -__ __ ____ '--"o Cooldown~~Rates i"= 1000 F/Hr .......................u steady-stateo-6o1750 60 oo250 -.... ...........M inim um I- ......:.. ........ .. ... ...4.. ... .........SBoltup Templi ..! ' i-500F !0 50 100 150 200 250 300 350 400 450 500 550Moderator Temperature (Deg. F)Figure 2-2 Sequoyahi~nit 2 Reactor Coolant System Cooldown Limitations (CooldownRates up to 100&deg;F/hr) Applicable for the First 32 EFPY (w/ Margins forInstrumentation Error of 100F and 60 psig) (Plotted Data provided on Table 2-2)2 PRESSURE TEMPERATURE LIMITS REPORTTable 2-1Sequoyah Unit 2 Heatup Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 10&deg;F and 60 psig)100 Heatup 1100 Critical Limit ILeak Test LimitT P T P JT P50 0 214 0 198 200050 591 214 607 214 248555 595 214 61460 601 214 62265 607 214 65770 614 214 65075 622 214 64780 630 214 64685 640 214 64890 646 214 *65395 646 214 661100 6.46 214 671105 646 214 680110 646 214 685115 646 214 701120 646 214 720125 648 214 743130 653 214 769135 661 214 798140 671 215 832145 685 220 869150 701 225 911155 720 230 959160. 743 235 1011165 769 240 1069170 798 245 1134175 832 250 1206180 869 255 1286185 911 260 1374190 959 265 1471195 1011 270 1579200 1069 275 1698205 1134 280 1829210 1206 285 1974215 1286 290 2134220 1374 295 2311225 1471230 1579235 1698240 1829245 1974250 2134255 23113 PRESSURE TEMPERATURE LIMITS REPORTTable 2-2Sequoyah Unit 2 Cooldown Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 1 0&deg;F and 60 psig)Steady State 1 20F 40F 60F I00OFT P T P T P T P T P5050556065707580859095100105110115120125130135140145*1501551601651761751801851901952002052102152202252300591595601607614622630640650661674688703720739.7607838098378689029409821028108011361199126813441429152216251739186520042158232850505560657075808590951001051101151201251301351401451501551601650552554558564572580589599610623636652668687707730755783814848885927.973102450505560657075808590951001051101151201251301351401451501551600503508514521529538548*55957158459961663465467670172975979383187291896850505560657075808590951001051101151201251301351401451501550.461466470478486496506518531546562580600622647674704738775816862913505055606570758085909510010511011512012513013514014515003663723803893994104234374534704905125365635936266637047498008564 PRESSURE TEMPERATURE LIMITS REPORTTable 3-1Selected Setpoints, Sequoyah Unit 2PORV#2 PORV#1Trcs (Deg.F) Setpoint (psig) Setpoint (psig)50 510 485100 580 555135 640 610174 745 682200 745 685*250 745 *685-278 745 685400 745 685450 2350 23505 PRESSURE TEMPERATURE LIMITS REPORTSequoyah Unit 2 LTOPS Selected Setpoints-/! [4-2500 --....J Lo- I I !0 01015 0 5 303040 5 0Reacto Coln Iyte enpraue(FA F--*-- FORV#2-FinaI --U-- FORV#1Figure 3-1 Sequoyah Unit 2 Selected LTOPS Setpoints (Plotted Data pro vided on Table 3-1)6 PRESSURE TEMPERATURE LIMITS REPORTTable 4-1Sequoyah Unit 2 Reactor Vessel Surveillance Capsule Withdrawal Schedule(a) Updated in Capsule Y dosimetry analysis (WCAP-15320t71).(b) Effective Full Power Years (EFPY) from plant startup.(c) Plant specific evaluation.(d) This fluence is not less than once or greater than twi ce the peak end of license (32 EFPY) fluence(e) Capsules 5, V, W and Z will reach a fluence of 2.71 x 1019 (EB> 1.0 MeV), the 48 EFPY peakvessel fluence at approximately 44 EFPY.,Administrative Note -The surveillance capsule withdrawal schedule in Table 4-1 is based on thesurveillance program for the original 40 year service life. Relocation of select standby, capsules toincrease the fluence lead factor in anticipation of the updated surveillance program for the 60 year licenserenewal service life is described in a TVA letter to NRC dated January 10, 2013 (ML13032A25 1). .Regulatory approval for the anticipatory standby capsule relocation has been granted by a NRC letter toTVA dated September 27, 2013 (ML13240A320). A complete update of the reactor vessel surveillanceprogram for the 60 year license renewal service life will be documented by a subsequent revision to thePTLR prior to entry into the license renewal extended operating period.7 PRESSURE TEMPERATURE LIMITS REPORTTable 5-1Comparison of the Sequoyah Unit 2 Surveillance. Material 30 ft-lb Transition Temperature Shifts andUpper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions30 ft-lb Transition Upper Shelf EnergyTemperature Shift DecreaseMaterial Capsule Fluence ' Predicted Measured Predicted Measured(x 1019 n/cm2) (OF)(a) (OF)(b) (%,/)(al) (%)(c)Intermediate Shell T 0.261 60.33. 63.65 17 12Forging 05 U 0.692 85.22 79.31 21 16(Tangential)(et285/X 1.22 100.23 85.7 23 8981057) Y 2.14 114.67 134.12 26 22Intermediate Shell T 0.261 60.33 48.73 17 7Forging 05 U 0.692 85.22 66.06 21 9(Axial)(et285/X 1.22 100.23 110.04 23 2981057) Y 2.14 .114.67 89.21 26 22Weld Metal T 0.261 43.12 74.56 20 2(Heat # 4278)(d) U 0.692 60.91 130.38 25 6X 1.22 71.63 44.22 29 35Y 2.14 81.96 86.91 33 "3HAZ Metal T 0.261 --24.58 -,- 2U 0.692 --64.03 --14X 1.22 --28.29 --19Y 2.14 --50.32 --39Notes:(a) Based on Regulatory Guide 1.99, Revision 2, methodology using the mean weight percent values ofcopper 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) Surveillance Weld was fabricated from weld wire type SMIT 89, Heat # 4278, Flux type SMIT 89,Lot # 1211.8 PRESSURE TEMPERATURE LIMITS REPORTTable 5-2Calculation of Chemistry Factors using Sequoyah Unit 2 Surveillance Capsule DataMaterial Capsule Capsule f(a) FF(b) ARTNDT(C) FF*ARTNDT FF2Intermediate Shell T 2.61E+18 0.635 63.7 40.45 0.403Forging 05 U 6.92E+18 0.897 79.3 71.13 0.805(Tangential) X 1.22E+19 1.055 85.7 90.41 1.113(Heat #288757 / Y 2.14E+19 1.207 134.1 161.86 1.457981057)Intermediate Shell T 2.61E+18 0.635 48.7 30.92 0.403Forging 05 U 6.92E+18 0.897 66.1 -59.29 0.805(Axial) X 1.22E+19 1.055 110.0 116.05 1.113(Heat #288757 / Y 2.14E+19 1.207 89.2 107.66 1.457981057) ______ ______SUM: 677.77&deg;F 7.556CFo5 = X(FF
* RTNDT) + X.( FF2) = (677.77) +(7.556) = 89.70FSurveillance Weld T 2.6 1E+18 0.635 69.4 (74.6) 44.07 0.403Material(d) *U 6.92E+18 0.897 121.3 (130.4) 108.81 0.805(Heat # 4278)(e) X 1.22E+19 1.055 41.1 (44.2) 43.36 1.113Y 2.14E+19 1.207 80.8 (86.9) 97.53 1.457SUM: 293 .77&deg;F 3.778CF Surv. weld =
* RTNDT) + X( FF2) = (293.77&deg;F) +(3.778) = 77.8"FNotes:(a) f = Calculated fluence from Capsule Y dosimetry analysis results [7], (n/cma2, E > 1.0 MeV).(b) FF = fluence factor = f02-.~o 3(c) AIRTNDT values are the measured 30 ft-lb shift values taken from App. B of Ref. 7, rounded to onedecimal point.(d) The surveillance weld metal ARTNDT Values have been adjusted by a ratio factor of 0.93.(e) Surveillance Weld was fabricated from weld wire type SMIT 89, Heat # 4278, Flux type SMIT 89, Lot #1211.
PRESSURE TEMPERATURE LIMITS REPORTTable 5-3Reactor Vessel Beltline Material Unirradiated TouglmessProperties for Sequoyah Unit 2Material Description Cu (%) Ni (%) Initial RTNDT(a)Intermediate Shell Forging 05(Heat #288757 /981057)013.710FLower Shell Forging 040.14 0.76 -22 &deg;F(Heat # 990469 / 293323)Intermediate to Lower Shell ForgingCircumferential Weld Seam W05(b) 0.12 0.11 -40F(Heat # 4278)Surveillance Weld(b) 0.13 0.11Notes:(a) The Initial RTNDT values are measured values(b) Circumferential Weld Seam was fabricated with weld wire type SMIT 89, Heat # 4278, Flux type SMIT89, Lot # 1211 and is representative of the intermediate to lower shell circumferential weld.10 PRESSURE TEMPERATURE LIMITS REPORTTable 5-4Peak Neutron Fluence Projections at Key Azimuthal Locations on the Reactor Vessel Clad/Base MetalInterface for Sequoyah Unit 2 (x 1019 n/cm2, E > 1.0 MeV)Azimuthal LocationEFPY 0&deg; 150 3 00 45010.54 0.211 0.336 0.426 0.63720 0.38 0.60 0.773 1.1632 0.593 0.934 1.21 1.8248 0.878 1.38 1.80 2.7111 PRESSURE TEMPERATURE LIMITS REPORTTable 5-5Sequoyah Unit 2 Calculation of the ART Values for the 1/4T Location @ 32 EFPY(a)Material RG 1.99 cF FF IRTNDT(b) ARTNDTo(C) Margin(d) ART(e)R2 Method (0F) _____ (0F) (0F) (0F) (0F)Position 1.1 95 1.027 10 97.6 34 142Intermediate Shell Forging 05Position 2.1 89.7 1.027 10 92.1 34 136Lower Shell Forging 04 Position 1.1 104 1.027 -22 106.8 34(0 119Intermediate to Lower Shell " Position 1.1 .63 '1.027 -4 64.7 56 117Circumferential Weld Seam Position 2.1 77.8 " 1.027 .-4 79.9 56(0 132Notes:(a) Neutron fluence value used for all materials is the highest value from Table 5-4 for 32 EFPY.(b) Initial RTNDoT values are measured values.(c) ARTrmT CF *FF(d) M =2 *(a'i2 +I 0'A2)1/2(e) ART =Initial RTNDT +- ARTNDTr + Margin (0F)(f) Data deemed not-credible (See Reference 7a), thus the full ca will be used to determine margin.Table 5-6Sequoyah Unit 2 Calculation of the ART Values for the 3/4T Location @ 32 EFPY(a)Material RG 1.99 CF FF IRTNDT(b) Margin(d) ART(e)R2 Method (0F) (0F) (0F) (0F) (0F)Position 1.1 95 0.745 10 70.8 34
* 115Intermediate Shell Forging 05Position 2.1 89.7 0.745 10 66.8 34 .111Lower Shell Forging 04 Position 1.1 104 0.745 -22 77.5 34(0 90intermediate to Lower Shell Position 1.1 63 0.745 .-4 46.9 56 99Circumferential Weld Seam Position 2.1 77.8 0.745 -4 58.0 56(0 110Notes: .(a) Neutron fluence value used for all materials is the highest value from Table 5-4 for 32 EFPY.(b) Initial RTrNDT values are measured values.(c) =CF *FF(d) M = 2 *(0i2 + raA)2(e) ART =Initial RTrNDv + ARTNDT + Margin (0F)(f) Data deemed not-credible (See Reference 7a), thus the full GA will be used to determine margin.12 PRESSURE TEMPERATURE LIMITS REPORTTable 5-7Summary of the Sequoyah Unit 2 Reactor Vessel Beitline Material ART ValuesMaterial RG 1.99 R2 1/4 ART 3/4 ARTMethod (0F) (0F)Position 1.1 142 115Intermediate Shell Forging 05Position 2.1 136 111Lower Shell Forging 04 Position 1.1 119 90Intermediate to Lower Shell Position 1.1 117 99Circumferential Weld Seam Position 2.1I 132 110Table 5-8RTPTs Calculations for Sequoyah Unit 2 BeltlineRegion Materials at 32 EFPY(a)Material Fluence FF CF ARTpTs~b) Margin RTNDT(U-)(C) RTpTs(d)(x 10 19 n/cm2, (0F) (0F) (0F) (0F) (0F)E>l.0 MeV)Intermediate Shell Forging 05 1.82 1.164 95 110.6 34 10 155Intermediate Shell Forging 05 1.82 1.164 89.7 104.4 34 10 148(Using S/C Data) _______Lower Shell Forging 04 1.82 .1.164 104 121.1 34(e) -22 133Circumferential Weld Metal 1.82 1.164 63 73.3 56 -4 125Circumferential Weld Metal 1.82 1.164 77.8 90.6 56(e) -4 143(Using S/C Data)____Notes:(a) Neutron fluence value used for all materials is the highest value from Table 5-4 for 32 EFPY.(b) ARTpTs= CF
* FF(c) Initial RTNDT values are measured values(d) RTpTs = RTNDT(U) + ARTpTs + Margin (0F)(e) Data deemed not-credible (See Reference 7a), thus the full ra.. will be used to determine margin.13 PRESSURE TEMPERATURE LIMITS REPORT6.0 References1. WCAP-8513, Tennessee Valley Authority Sequoyah Unit No. 2 Reactor Vessel RadiationSurveillance Program, J. A. Davidson, et. al., November 1975.2. Code of Federal Regulations, 10OCFR50, Appendix H, Reactor Vessel Material SurveillanceProgr~am Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C.3. ASTM E23 Standard Test Method Notched Bar" Impact Testing of Metallic Materials, in ASTMStandards, American Society for Testing and Materials, Philadelphia, PA.4. Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G, Fracture ToughnessCriteria for Protection Against Failure5. ASTM E185-82, Annual Book of ASTM Standards, Section 12, Volume 12.02, Standard Practicefor Conducting Surveillance Tests for Light- Water Cooled Nuclear Power Reactor Vessels.6. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S.Nuclear Regulatory Commission, May 1988.7a. WCAP- 15320, Analysis of Capsule Y fi'omn the Tennessee Valley Authority Sequoyah Unit 2Reactor Vessel Radiation Surveillance Program, T.J. Laubham, et. al., November 1999.7b. WCAP-*10509, Analysis of Capsule T fiom the. Tennessee Valley Authority Sequoyah Unit 2Reactor Vessel Radiation Surveillance Program, R. S. Boggs, et al, April 1984.7c. Southwest Research Institute Nondestructive Evaluation Science and Technology Division,Reactor Vessel Material Surveillance Program and Technology Division, Reactor Vessel Material"Surveillance Pro gram for Sequoyah Unit 2." Analysis of Capsule U, Final Report SwRI Project*17-8851 TVA Contra~ct 85PJH-964430, January !1990.7d. WCAP- 13545, Analysis of Capsule X from the Tennessee Valley Authority Sequoyah Unit 2Reactor Vessel Radiation Surveillance Program, M. A. Ramirez, S. L. Anderson, A. Madeyski,November 1992.8. CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.1, developed by ATIConsulting, March 1999.9. WCAP- 14040-NP-A, Revision 2, Methodology Used to Develop Cold Overpressure MitigatingSystem Setpoints and RCS Heatup and Cooldown Limit Curves, J.D. Andrachek, et. al., January1996.10. WCAP-15321, Revision 2, Sequoyah Unit 2 Heatup and Cooldown Limit Curves for NormalOperation and PlTLR Support Documentation, J.H. Ledger, et.al., July 2003.11. ASME Code Case N-640, Alternative Reference Fracture Toughness for Development of P-T*Limit Curves for Section XL, Division 1, dated February 26, 1999.14 PRESSURE TEMPERATURE LIMITS REPORT12. WCAP-15984-P, Revision 01, Reactor Vessel Closure Head/Vessel Flange RequiremnentsEvaluation For Sequoyah Units 1 and 2, W. Bamford, et.al., April 2003.13. Westinghouse Letter to TVA, TVA-93-105, Cold Overpressure Mitigation System Code Case andDelta-P Calculation, dated May 19, 1993.14. Calculation SQN-IC-0 14, Demonstrated Accuracy Calculation for Cold Overpressure ProtectionSystem.15 Tennessee Valley Authority, Post Office Box 2000, Soddy Daisy, Tennessee 37384-2000November 13, 201510 CFR 50.4ATT-N: Document Control DeskU.S. Nuclear Regulatory CommissionWashington, D.C. 20555-0001Sequoyah Nuclear Plant, Units I and 2Renewed Facility Operating License Nos. DPR-77 and DPR-79NRC Docket Nos. 50-327 and 50-328
 
==Subject:==
Sequoyah Unit I Pressure Temperature Limits Report,Revision 5, and Sequoyah Unit 2 Pressure Temperature LimitsReport, Revision 6
 
==References:==
: 1. Letter from NRC to TVA, "Sequoyah Nuclear Plant, Units I and 2-Issuance of Amendments for the Conversion to the ImprovedTechnical Specifications with Beyond Scope Issues (TAC Nos.MF3128 and MF3129)," dated September 30, 20015(MLI15238B460)In accordance with Sequoyah Nuclear Plant (SQN) Units 1 and 2 TechnicalSpecifications (Tss) 5.6.4.c, enclosed is the Unit I Pressure Temperature LimitsReport (PTLR), Revision 5, and Unit 2 PTLR, Revision 6. In accordance with TSs5.6.4.c, the PTLRs are required to be provided to the Nuclear RegulatoryCommission (NRC) within 30 days after any revision. Sequoyah Units I and 2 wereissued license amendment Nos. 334 and 327, respectively for improved standardTSs (Reference 1). These license amendments resulted in the enclosed revisions toeach of the PTLRs. The revisions also include other editorial clarifications andadministrative changes identified during the revision process as described inEnclosure 1. The revised PTLRs became effective on October 16, 2016.There are no new regulatory commitments in this letter. If you have any questions,please contact Jonathan Johnson, SQN Site Licensing Manager at (423) 843-8129.
U.S. Nuclear Regulatory CommissionPage 2November 13, 2015Sequo Nuclear PlantEnclosures1.2.3.Units 1 and 2 Pressure Temperature Limits Report ChangesSequoyah Unit 1 Pressure Temperature Limits Report, Revision 5Sequoyah Unit 2 Pressure Temperature Limits Report, Revision 6ZTK: DVGEnclosurescc (Enclosures):NRC Regional Administrator -Region IINRC Senior Resident Inspector -SQN ENCLOSURE1ISEQUOYAH UNITS 1 AND 2PRESSURE TEMPERATURE LIMITS REPORT CHANGESThe following describes the editorial clarifications and administrative changes madeto each Units' Pressure Temperature Limits Report (PTLR).1. Conflict between the figure index referencing Cold Overpressure MitigationSystem (COMS) and the figure title referencing Low TemperatureOverpressure Protection System (LTOPS) was Corrected for consistency withthe system terminology in Section 3.4.12 of the Technical Specifications.2. Section 1, "RCS Pressure Temperature Limits Report (PTLR)," was revisedto clarify which Limiting Condition for Operations are affected by the PTLR.3. Section 3.1, "Pressurizer PORV Lift Setting Limits," was revised forconsistency with the analysis of record contained in Topical Report No.WCAP-1 5293, Revision 2 "Sequoyah Unit 1 Heatup and Cooldown LimitCurves for Normal Operation and PTLR Support Documentation," andTopical Report No. WCAP-15321 Revision 2 "Sequoyah Unit 2 Heatup andCooldown Limit Curves for Normal Operation and PTLR SupportDocumentation".4. Section 5.0, "Supplemental Data Tables," added the word "Forging" to thesentence in refer to the limiting beltline material identified in Tables 5-5 and5-6.5 A note was added to Table 4-1, "Sequoyah Unit 1 Reactor VesselSurveillance Capsule Withdrawal Schedule," to describe current and futureactivity with the surveillance capsule withdrawal schedule.6. Tables 5-5 and 5-6 were revised with a different annotation conventionconsistent with the other tables in the PTLR.
ENCLOSURE 2SEQUOYAH UNIT 1PRESSURE TEMPERATURE LIMITS REPORT, REVISION 5 B88 15 t 16800MAPRESSURE TEMPERATURE LIMITS REPORT....APPROVEDThis qp~provaIl does. not reliev, theContractor from any port o9 his re-,*ponsibility Jar, the correctness of.* design. debugailnd dimlnsio~ns.YLetter lo.Nl107l6.*oot,: ,October 16, 2015. -.SQEP BY: W. JX PierceTennessee Valley AuthoritySequoyah unit 1Pressure Temperature Limits ReportRevision 5, September 2015PROJECT Sequovah .DISCIPLINE N"CONTRACT .4411 1JNT' 1DESC. RCS Pressure-Temperature Limit Report*DWG/DOC NO. PTLR-1SHEET -OF -. "REV. 05*DATE 10/16/15 ECN/DCN -. FILE N2N-081EDMS, WT CA-K PRESSURE TEMPERATURE LIMITS REPORTTable of ContentsList of Tables .................................................................................................... ivList of Figures......................................................................................................v1.0 RCS Pressure Temperature Limits Report (PTLR) ......................................... ....... 12.0 Operating Limits.......................................................................................... 12.1 RCS Pressure/Temperature (PIT) Limits (TS 3.4.3) ............................................... 13.0 Low Temperature Overpressure Protection System (TS 3.4.12)...... ............................ 13.1 Pressurizer PORV Lift Setting Limits .............................................................. 23.2 Arming Temperature................................................................................ 24.0 Reactor Vessel Material Surveillance Program...................................................... 25.0 Supplemental Data Tables............................................................................... 36.0 References................................................................................................ 19ii PRESSURE TEMPERATURE LIMITS REPORTList of TablesTable 2-1 Sequoyah Unit 1 Heatup Limits at 32 EPPY(with Uncertainties for Instrumentation Errors of 10&deg;F and 60 psig)......... ...................... 6Table 2-2 Sequoyah Unit 1 Cooldown Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 1 0&deg;F and 60 psig)................................STable 3-1 Selected Setpoints, Sequoyah Unit 1.......................................................... 10Table 4-1 Sequoyah Unit I Reactor Vessel Surveillance Capsule Withdrawal Schedule ............ 12Table 5-1 Comparison of the Sequoyah Unit 1 Surveillance Material 30 ft-lb TransitionTemperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99,revision 2, Predictions .............................'............................................. 13Table 5-2 Calculation of Chemistry Factors using Sequoyah Unit 1 Surveillance Capsule Data. .... 14Table 5-3 Reactor Vessel Beltline Material Unirradiated Toughness Properties forSequoyah Unit 1 ............................................................................ ..... 15Table 5-4 Peak Neutron Fluence Projections at Key Azimuthal Locations on the Reactor VesselClad/Base Metal Interface for Sequoyah Unit 1 (x 10'9 n/cm2, B > 1.0 MeV) ............ 16Table 5-5 Sequoyah Unit 1 Calculation of the ART Values for the 1 /4T Location @ 32 EFPY.....17Table 5-6 Sequoyah Unit 1 Calculation of the ART Values for the 3/4T Location @ 32 EFPY ......17Table 5-7 Summary of the Limiting ART Values Used in the Generation of the Sequoyah Unit 1Heatup/Cooldo~vn Curves. .................................................................... 18Table 5-8 RTm-s Calculatiohis for Sequoyah Unit 1 Beitline Region Materials at 32 EFPY.......... 18iii PRESSURE TEMPERATURE LIMITS REPORTList of FiguresFigure 2-1 Sequoyah Unit 1 Reactor Coolant System Heatup Limitations(Heatup Rate of 1 00&deg;F/hr) Applicable for the First 32 EFPY(w/ Margins for Instrumentation Errors of 1 0&deg;F and 60 psig)................................ 4Figure 2-2 Sequoyah Unit 1 Reactor Coolant System Cooldown Limitations(Cooldown Rates up to 100&deg;F/hr) Applicable for the First 32 EFPY(w/ Margins for Instrumentation Errors of 1 0&deg;F and 60 psig)................................ 5Figure 3-1 Sequoyah Unit 1 Selected LTOPS Setpoints ................................................. 11iv PRESSURE TEMPERATURE LIMITS REPORT1.0 RCS Pressure Temperature Limits Report (PTLR)This PTLR for Sequoyah Unit 1 has been prepared in accordance with the requirements of TecirnicalSpecification (TS) 5.6.4. Revisions to the PTLR shall be provided to the NRC after issuance.This report affects TS 3.4.3, RCS Pressure/Temperature Limits (PiT) Limits and TS 3.4.12, LowTemperature Over Pressure Protection (LTOP) System.2.0 RCS Pressure and Temperature LimitsThe limits for TS 3.4.3 are presented in the subsections which follow and were developed using the NRCapproved methodologies specified in TS 5.6.4 with exception of ASME Code Case N-640[13] (Use of Kit),WCAP- 15984-P['14 (Elimination of the Flange Requirement), 1996 Version of Appendix Gm and the-revised fluencesE7T. The operability requirements associated with LTOPS are specified in TS 3.4.12 andwere determined to adequately protect the RCS against brittle fracture in the event of an LTOP Transientin accordance with the methodology specified in TS 5.6.4.2.1 RCS Pressure/Temperature (PIT) Limits (TS 3.4.3)2.1.1 The minimum boltup temperature is 500F2.1.2 The RCS temperature rate-of-change limits are:a. A maximum heatup rate of 1 00&deg;F in any one hour period.* b. A maximum cooldown rate of l00OF in any one hour period.c. A maximum temperature change of less than or equal to I10&deg;F in any one hour period duringinservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.2.1.3 The RCS P/T limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticalityare specified by Figures 2-1 and 2-2.3.0 Low Temperature Overpressure Protection System (TS 3.4.12)The lift setpoints for the pressurizer Power Operated Relief Valves (P.ORVs) are presented in thesubsections which follow. These lift setpoinlts have been developed using the NRC-approvedmethodologies specified in TS 5.6.4.1 PRESSURE TEMPERATURE LIMITS REPORT3.1 Pressurizer PORV Lift Setting LimitsThe pressurizer PORV lift setpoints are specified by Figure 3-1 and Table 3-1 (Ref. 10). The limits for theLTOPS setpoints are contained in the 32 EFPY steady-state curves (Table 2-2), which are beitlineconditions and are not compensated for pressure differences between the pressurizer transmitter and thereactor midplane/beltline or for instrument inaccuracies. The pressure difference between the pressurizertransmitter and the reactor vessel midplane/beltline with four reactor coolant pumps in operation is 68.3psi (Ref. 11).Note: These setpoints include allowance for the pressure difference between the pressurizertransmitter and the reactor vessel midplane/beltline and the SOTF thermal transport effect forheat injection transients. A demonstrated accuracy calculation (Reference 12) has beenperformed to confirn that the setpoints will maintain the system pressure within theestablished limits when the pressure difference between the pressure transmitter and reactormidplane and maximum temperature/pressure instrument uncertainties are applied to thesetpoints.3.2 Anning TemperatureThe LTOPS arming temperature is based upon the methodology defined in the Sequoyah Nuclear PlantUnit 1 Technical Specifications Administrative Controls Section 5.6.4. The arming temperature shall be _<350&deg;F.4.0 Reactor Vessel Material Surveillance ProgramThe reactor vessel material irradiation surveillance specimens shall be removed and examined todetermine changes in material properties. The removal schedule is provided in Table 4-1. The results ofthese examinations shall be used to update Figures 2-1, 2-2 and 3-1.The pressure vessel steel surveillance program (WCAP-8233 r1l) is in compliance with Appendix H to 10CFR 50, "Reactor Vessel Material Surveillance Program Requirementsr21."' The material test requirementsand the acceptance standard utilize the reference nil-ductility temperature which is determined inaccordance with ASTM E23 [3]. The empirical relationship between RTNDr and the fracture toughness ofthe reactor vessel steel is developed in accordance with Code Case N-640 of Section XI of the ASMEBoiler and Pressure Vessel Code, Appendix 0, "Fracture Toughness Criteria for Protection AgainstFailureE41.'' The surveillance capsule removal schedule meets the requirements of ASTM E185-82N5.Theremoval schedule is provided in Table 4-1.2 PRESSURE TEMPERATURE LIMITS REPORT5.0 Supplemental Data TablesTable 5-1 contains a comparison of measured surveillance material 30 ft-lb transition temperature shiftsand upper shelf energy decreases with Regulatory Guide 1.99, Revision 2[6], predictions.Table 5-2 shows calculations of the surveillance material chemistry factors using surveillance capsuledata. Note that in the calcuilation of the surveillance weld chemistry factor, the ratio procedure fromRegulatory Guide 1.99, Revision 2 was followed. The ratio in question is equal to 0.90.Table 5-3 provides the required Sequoyah Unit 1 reactor vessel toughness data.Table 5-4 provides a summary of the fluence values used in the generation of the heatup and cooldownlimit curves and the PTS evaluation.Table 5-5 and 5-6 show the calculation of the 1/4T and 3/4T adjusted reference temperature at 32 EFPYfor each beitline material in the Sequoyah Unit 1 reactor vessel. The limiting beitline material was theLower Shell Forging 04.Table 5-7 provides a summary of the adjusted reference temperature (ART) Values of the Sequoyah Unit 1reactor vessel beltline materials at the 1/4T and 3/4T locations for 32 EFPY.Table 5-8Sprovides RTPTs values for Sequoyah Unit 1 at 32 EFPY.3 PRESSURE TEMPERATURE LIMITS REPORTMATERIAL PROPERTY BASISLTMITING MATERIAL: LOWER SHELL FORGING 04LIMITING ART VALUES AT 32 EFPY: 1/4T, 216&deg;F3/4T, 186&deg;F2500 --_o- erlim Version.5.1 Run.15680 [i /-2250i Leak Test Limit /-; -2000 .......Unacceptable ....... ... -- --"Acceptable ...,Operation, Operation__:Heatup Rate Critical Limit100o Deg. F/HrI 100 Deg. F/ HrI-- "~2=i(n 'i.120 ,.. .. ....S 1000 --- _ __- ___ __ _750 -- -- -__ _ _ -__ ___-__ __ __ __ --- __/ Criticality Limit based on Iinservice hydrostatic test500 -temnperatuire (2880F) for the __service period up to 32 EFPY(IMinimum !250 ......-~ Boltup .. ... -; -___ __) _STemp = 50*F0 50 100 150 200 250 300 350 400 450 500 550Moderator Temperature (Deg. F)Figure 2-1 Sequoyah Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of100&deg;F/hr)-Applicable for the First 32 EFPY (w/Margins for Instrumentation Errorof 10&deg;F and 60 psig) (Plotted Data provided on Table 2-1)4 PRESSURE TEMPERATURE LIMITS REPORTMATERIAL PROPERTY BASISLIMITING MATERIAL: LOWER SHELL FORGING 04LIMITING ART VALUES AT 32 EFPY: 114T, 216&deg;F3/4T, 186&deg;F2500225020001750150021250o 10007505002500 50 100 150 200 250300 350 400 450 500 550Moderator Temperature (Deg. F)Figure 2-2 Sequoyah Unit 1 Reactor Coolant System Cooldown Limitations (CooldownRates up to 100&deg;F/hr) Applicable for the First 32 EFPY (w/Margins forInstrumentation Error of 10&deg;F and 60 psig) (Plotted Data provided on Table 2-2)5 PRESSURE TEMPERATURE LIMITS REPORTTable 2-1Sequoyah Unit 1 Heatup Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 10&deg;F and 60 psig)100 Heatup 1100 Critica] Limit ILeak Test LimitT P J T Pj T P50 0 288 0 272 200050 477 *288 477 288 248555 477 288 47760 477 288 47765 477 288 47770 477 288 47875 477 288 47880 477 288 48085 477 288 48190 477 ,288 48395 477. 288 485100 477 288 487105 477 288 490110 477 288 493115 477 288 497120 477 288 500125 477 288 505"130 477 288 508135 477 288 515140 477 288 517145 .477 288 527150 477 288 528155 478 288 541160 480 288 541165 483 288 555170 487 288 557175 493 288 571180 500 288 575185 508 288 589190 517 288 609195 528 288 631200 541 288 .656205 555 ,288 684210 571 288 714215 589 .288 748220 609 290 786225 631 295 8286 PRESSURE TEMPERATURE LIMITS REPORTTable 2-1 -(Continued)Sequoyah Unit 1 Heatup Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 1 0&deg;F and 60 psig)100 Heatup 100 Critical LimitT P T P230 656 300 874235 684 305 925240 714 310 981245 748 315 1044250 786 320 1112255 828 325 1188260 874 330 1272265 925 335 1364270 981 340 1466275 1044 345 1578280 1112 350 1702285 1188 355 1838290 1272 360 1988295 1364 365 2154300 1466 370 2337305 1578310 1702315 1838320 1988325 2154330 23377 PRESSURE TEMPERATURE LIMITS REPORTTable 2-2Sequoyah Unit 1 Cooldown Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 10&deg;F and 60 psig)Steady State [20F )40F P1 60F 100F50505560657075808590951001051101151201251301351401451501551601651701751801851901952002052102150552553555556558560561564566569571575578582586591596602608616623632642652664677691707724743764788814843505055606570758085,909510010511011512012513013514014515015516016517017518018519019520020521021505035055075095105125145165185215245275315355405455505565635715795885996106236376526696887097337597878195050556065707580859095100105110115120125130135140145150155160165170175180185190195200205210215045745845946046246446546847047347647948348749249750351051752553454455656858259761463365467770273176279750505560657075808590951001051101151201251301351401451501551601651701751801851901952002052102150408409410411412414416418420423426430434.4384434494564634714794895005125265415585775976206466747057407795050556065707580859095100105110115120125130'1351401451501551601651701751801851901952002052102150305306307308309311.313315318.321325329333338344351358367376"387399412427443461482505530558590624663706754______________________________________ +/- _______________________________________ _________________________________________ I8 PRESSURE TEMPERATURE LIMITS REPORTTable 2-2 -(Continued)Sequoy~a Unit 1 Cooldown Limits at 32 EFPY(without Uncertainties for Instrumentation Errors)Steady State 20F 40F 60F 100F.T P T P T PT P T P220 874 220 853 220 836 220 821 220 806225 909 225 892 225 878 225 869 2-25 865230 948 230 935 230 925 230 921235 991 235 982 235 978240 1038 240 1034245 1090250 1148255 1212260 1283265 1360270 1447275 1542280 1647285 1763290 1892295 2034300 2191305 23649 PRESSURE TEMPERATURE LIMITS REPORT*Table 3-1Selected Setpoints, Sequoyah Unit 1Trs Dg.) PORV#2 PORV#1Trcs(De.F)Setpoint (psig) Setpoint (psig)50 490 465100 500 475135 540 510175 575 540200 610 570250 745 685280 745 685405 745 685450 2350 235010 PRESSURE TEMPERATURE LIMITS REPORTSequoyah Unit I LTOPS Selected Setpoints250020002i1500100005000 50 100 150 200 250 300 350 400 450 500Reactor Coolant System Temnperature (&deg;F)--4-- FOFRV#2 Setpoint --=- PORV#1 SetpointFigure 3-1 Sequoyah Unit 1 Selected LTOPS Setpoints (Plotted Data provided on Table 3-1)11 PRESSURE TEMPERATURE LIMITS REPORTTable 4-1Seqtuoyah Unit 1 Reactor Vessel Surveillance Capsule Withdrawal Schedule* Removal Time FluenceCapsule Location Lead Factorca) (EFPY) (b) (n/cm2,E>l.0 1T 400 3.39 1.03 2.61 x 10O8 (c)U 1400 3.47 3.00 7.96 x 1018 (c)X 2200 3.47 5.27 1.32 x 1019 (c)Y 320&deg; 3.43 10.03 2.19 x 10'9 (c,d)5 40 1.08 Standby (d,e)V 176&deg; 1.08 Standby (d,e)W 184&deg; 1.08 Standby (d,e)Z 3560 1.08 Standby (d,e)Notes:(a)(b)(c).(d)(e)Updated in Capsule Y dosimetry analysis Effective Full Power Years (EFPY) from plant startup.Plant specific evaluation.This fluence is not less than once or greater than twice the peak end of license (32 EFPY) fluenceCapsules 5, V, W and Z will reach a fluence of 2.74 x 1019 (E > 1.0 Mev), the 48 EFPY peakvessel fluence at approximately 44 EFPY, respectively.Administrative Note -The surveillance capsule withdrawal schedule in Table 4-1 is based on thesurveillance program for the original 40 year service life. Relocation of select standby capsules toincrease the fluence lead factor in anticipation of the updated surveillance program for the 60 year licenserenewal service life is described in a TVA letter to NRC dated May 14, 2015 (ML15!34A377).Regulatory approval for the anticipatorY standby capsule relocation has been granted by a NRC letter toTVA dated September 4, 2015 (ML 15244B222). A complete update of the reactor vessel surveillanceprogram for the 60 year license renewal service life will be documented by a subsequent revision to thePTLR prior to entry into the license renewal extended operating period.12 PRESSURE TEMPERATURE LIMITS REPORTTable 5-1Comparison of the Sequoyah Unit 1 Surveillance Material 3 0 ft-lb Transition Temperature Shifts andUpper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions30 ft-lb Transition Upper Shelf EnergyTemperature Shift DecreaseMaterial Capsule Fluence Predicted Measured Predicted Measured____________ ______ (X 1019 n/cm2). (OF)(a) (OF)(b) (%)(a) .(%)(C)Lower Shell T 0.261 .59.85 67.52 16 16Forging 04 U 0.796 89.3 109.7 20.5 21(Tneta)X 1.32 102.6 145.12 23 8(Heat # 980919 /281587) Y 2.19 114.95 129.87 26.5 23Lower Shell T 0.261 59.85 50.59 16 0Forging 04 U 0.796 89.3 67.59 20.5 19(xa)X 1.32 102.6 103.34 23 22(Heat # 980919 / _____281587) Y 2.19 114.95 133.35 26.5 19Weld Metal T 0.261 111.13 127.79 .35 30(Heat # 25295)(d) U 0.796 165.82 144.92 42 26X 1.32 190.51 159.02 .45 .21Y 2.19 213.44 163.8 48 28IiAZ Metal T 0.261 --45.48 --20U 0.796 --. 78.94 --26X 1.32 --95.89 --3Y 2.19 --73.3 --10Notes:(a) Based on Regulatory Guide 1.99, Revision 2, methodology using the mean weight percent values ofcopper and nickel of the surveillance material.(b) .Calculated using measured Charpy data plotted using CVGRAPH, Version 4.1 [8(c) Values are based on the definition of upper shelf energy given in ASTM E185-82.(d) Surveillance Weld was fabricated from weld wire type SMIT 40, Heat # 25295, Flux type SMIT 89,Lot # 1103.13 PRESSURE TEMPERATURE LIMITS REPORTTable 5-2Calculation of Chemistry Factors using Sequoyah Unit 1 Surveillance Capsule DataMaterial Capsule Capsule f~a) FF) ARTNBT(C) FF*ARTNDT FF2Lower Shell T 2.61E+18 0.63 67.52&deg;F 42.54&deg;F 0.40Forging 04 U 7.96E+18 0.94 109.7&deg;F 103.12&deg;F 0.88(Tangential)X 1.32E+19 1.08. 145.12&deg;F 156.73&deg;F 1.16(Heat # 980919/ /_________281587) Y 2.19E+19 1.21 129.87&deg;F 157.14&deg;F 1.47Lower Shell T 2.6 1E+I18 0.63 50.59&deg;F 31.87&deg;F 0.40Forging 04 U 7.96E+18 0.94 67.59&deg;F 63.53&deg;F 0.88(Axial) X 1.32E+19 1.08 103.34&deg;F 111.61&deg;F 1.16(Heat #980919 / *Y 2.19E+19 1.21 133.35&deg;F 161.35&deg;F 1.47281587)SUM: 827.89&deg;F 7.82CF04 =
* RTNDTr) + 2( FF2) = (827.89) +(7.82) = 105.9&deg;FSurveillance Weld T 2.61E+18 0.63 115.0&deg;F 72.5&deg;F 0.40Material(d) U 7,96E+18 0.94 130.4&deg;F 122.6&deg;F 0.88(Heat # 25295)(e) X 1.32E+19 1.08 143.1&deg;F 154.5&deg;F 1.16Y 2.19E+19 1.21 147.4OF 178.40F 1.47SUM: 528.0&deg;F 3.91CF Surv. we~d =XY(FF
* RTrrDT) +X( FF2) =(528.0&deg;F) + (3.91) = 135.0&deg;FNotes:(a) f= Calculated fluence from Capsule Y dosimetry analysis resultsETI, (n/cm2, E > 1.0 MeV).(b) FF = fluence factor =fI0.28"0.1logf.(c) ARTNDTValues are the measured 30 ft-lb shift values taken from App. B of Ref. 7, rounded to onedecimal point.(d) The surveillance weld metal ARTNDoT values have been adjusted by a ratio factor of 0.90.(e) Surveillance Weld was fabricated from weld wire type SMIT 40, Heat # 25295, Flux type SMIT 89,Lotft 110314 PRESSURE TEMPERATURE LIMITS REPORTTable 5-3Reactor Vessel Beltline Material Unirradiated Toughness Properties for Sequoyah Unit 1Material Description Cu (%) Ni (%) Initial RTNDT(a)Intermediate Shell Forging 05(Heat'#980807/281489) 01 .640Lower Shell Forging 04(Heat #980919/281587) 01 .670Surveillance Weld (Heat # 25295)(b'd, e) == 0.387 0.11 ---Rotterdam Test(c. e) 0.30 ......-Rotterdam Test(c' e) 0.25 ......Rotterdam Test(c' e) 0.46 ......-Best Estimate of the Intermediate to Lower ShellForging Circumferential Weld Seam W05 0.35 0.11 -40OF(Heat # 25295)(d. e)(a) The Initial RTNDT values are measured values(b) These copper and nickel values are best estimate values for only the surveillance weld metal and is the averageof three data points [0.424 (WCAP-10340, Rev.1), 0.406 (WCAP-10340, Rev.1), 0.33 (WCAP-8233) copperand 0.084 (WCAP-10340, Rev.1), 0.085 (WCAP-10340, Rev.1), 0.17 (WCAP-8233) nickel.]. These values aretreated as one data point in the calculation of the best estimate average for the inter, to lower shell circ. weldshown above. Originally the 0.424 / 0.406 and 0.084 / 0.085 values were reported as single points, 0.41 -0.42and 0.08 (Per WCAP-10340, Rev. l[7d]), but it is actually made up of two data points. Sample TW58 fromCapsule T was broken into two samples, TW58a and TW58b, thus providing the two data points.(c) From NRC Reactor Vessel Integrity Database (RVID) and ultimately fr'om Rotterdam Weld Certifications.(d) Circumferential Weld Seam W05 was fabricated with weld wire type SMvIT 40, Heat # 25295, Flux type SMVIT89, Lot # 2275. The surveillance weld was fabricated with weld Wire type SMIT 40, Heat # 25295, Flux typeSMIT 89, Lot # 1103 and is representative of the intermediate to lower shell circumferential weld.(e) The surveillance weld and the three Rotterdam tests are averaged together for the Best Estimate of theIntermediate to Lower Shell Forging Circumferential Weld Seam.15 PRESSURE TEMPERATURE LIMITS REPORTTable 5-4Peak Neutron Fluence Projections at Key Azimuthal Locations on the Reactor Vessel Clad/Base MetalInterface for Sequoyah Unit 1 (x 10 9 n/cm2, E > 1.0 MeV)Azimuthal LocationEFPY 00 150 300 45010.03 0.205 0.321 0.409 0.63720 0.387 0.596 0.761 1.1832 0.605 0.928 1.19 1.8448 0.896 1.37 1.75 2.7216 PRESSURE TEMPERATURE LIMITS REPORTTable 5-5Sequoyah Unit 1 Calculation of the ART Values for the 1/4T Location @ 32 EFPY(a)Material RG 1:99 CF FF ARTnDT(C) Margin(d) ART(e)R2 Method (0F) (&deg;F) (0F) (0F) (0F)Intermediate Shell Forging 05 Position 1.1 115.6 1.029 40 119.0 34 193Position 1.1 95 1.029 73 97.8 34 205Lower Shell Forging 04Position 2.1 105.9 1.029 73 109.0 34(0 216Intermediate to Lower Shell Position 1.1 161.3 1.029 -40 166.0 56 182Circumferential Weld Seam Position 2.1 135.0 1.029 -40 138.9 56(0 155Notes:(a) Neutron fluence value used for all materials is the highest value from Table 5-4 for 32 EFPY.(b) Initial RTNIJT values are measured values.(c) AXRTNDT = CF
* FF(d) M = 2 *(a'2 +- OA2)1/2(e) ART = Initial RTNDT + ARTNDT + Margin (0F)(f) Data deemed not-credible (See Reference 7a), thus the full GA will be used to determine margin.Table 5-6Sequoyah Unit 1 Calculation of the ART Values for the 3/4T Location @ 32 EFPY(a)Material RG 1.99 CF FF IRTrmT(b) ART~rTC() Margin(d) ART(e)*R2 Method (0F) (0F) (0F) (0F) (0F)Intermediate Shell Forging 05 Position 1.1 115.6 0.747 40 86.4 34 160Loe hl ogn 4 Position 1.1 *95 0.747' 73 71.0 34 178*Position 2.1 105.9 0.747 73 79.1
* 34(0 186Intermediate to Lower Shell Position 1.1 161.3 0.747 -40 120.5 56 137Circumferential Weld Seam Position 2.1
* 135.0 0.747 -40 100.8 56(0 117Notes:(a) Neutron fluence value used for all materials is the highest value from Table 5-4 for 32 EFPY.(b) Initial RTNDT values are measured values.(c) ARTrDT =CF *FF(d) M = 2 *(a.2 + Gra2)1t2(e) ART = Initial RTNDT .+- ARTrNDT + Margin (0F)(f) Data deemed not-credible (See Reference 7a), thus the full GrA will be used to determine margin.17 PRESSURE TEMPERATURE LIMITS REPORTTable 5-7Summnary of the Sequoyah Unit 1 Reactor Vessel Beitline Material ART ValuesMaterial RG 1.99 R2 1/4 ART 3/4 ARTMethod (0F) (0F)Intermediate Shell Forging 05 Position 1.1 193 160Position 1.1 205 178Lower Shell Forging 04Position 2.1 216 186Intermediate to Lower Shell Position 1.1 182 137Circumferential Weld Seam Position 2.1 155 117Table 5-8RTP-s Calculations for Sequoyah Unit 1 Beltline .Region Materials at 32 EFPY(a)Material Fluence FF CF ARTpTs~b) Margin RTpxs(d)(X 1019 n/cm2, "(F) (0F) (0F) (OF) (0F)E>1.0 MeV)Intermediate Shell Forging 05 1.84 1.167 115.6 .134.9 34 40 209Lower Shell Forging 04 1.84 1.167 95.0 110.9 34 73 218Lower Shell Forging 04 1:84 1.167 105.9 123.6 34(e) 73 231(Using S/C Data)Circumferential Weld Metal 1.84 1.167 161.3 188:2 56 -40 204Circumferential Weld Metal 1.84 1.167 135.0 157.5 56(e) -40 174(Using S/C Data)Notes:(a) Neutron fluence value used for all materials is the highest value from Table 5-4 for 32 EFPY.(b) ARTPTs = CF *FF(c) Initial RTNOT values are measured values(d) RTPTs= + ARTpTs + Margin (0F)(e) Data deemed not-credible (See Reference 7a), thus the full will be used to determine margin.18 PRESSURE TEMPERATURE LIMITS REPORT6.0 References1. WCAP-8233, Tennessee Valley Authority Sequoyah Unit No. 1 Reactor Vessel RadiationSurveillance Progr~am, S. E. Yanichko, et. al., December 1973.2. Code of Federal Regulations, 10OCFR50, Appendix H, Reactor Vessel Material SurveillancePro gramn Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C..3. ASTM E23 Standard Test Method Notched Bar himpact Testing of Metallic Materials, in ASTMStandards, American Society for Testing and Materials, Philadelphia, PA.4. Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G, Fracture Toughness"Criteria for Protection Against Failure5. ASTM E185-82, Annual Book of ASTM Standards, Section 12, Volume 12.02, Standard Practicefor Conducting Surveillance Tests for Light- Water Cooled Nuclear Power Reactor Vessels.6. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S.Nuclear Regulatory Commission, May 1988.7a. WCAP-15224, Analysis of Capsule the Tennessee Valley Author"ity Sequoyah Unit 1Reactor Vessel Radiation Surveillance Program, T.J. Laubham, et. al., June 1999.7b. WCAP-1 3333, Analysis of Capsule Xfoino the Tennessee Valley Authority Sequoyah Unit 1Reactor Vessel Radiation Surveillance Progr~am, M.A. Ramirez, S. L. Anderson, L. Albertin, June1992.7c. .SwRI Project 06-8851, Reactor Vessel Material Surveillance Progr'am for Sequoyah Unit No. 1:Analysis of Capsule U, P. K. Nair, et al., October 1986.7d. WCAP- 10340, Revision 1, Analysis of Capsule T foin the Tennessee Valley Authority SequoyahUnit.] Reactor Vessel Radiation Surveillance Program, S.E. Yanichko, et. al., February 1984.8. CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.1, developed by ATIConsulting, March 1999.9. WCAP-14040-NP-A, Revision 2, Methodology Used to Develop Cold Overpressure MitigatingSystem Setpoints and RCS Heatup and Cooldown Limit Curves, J.D. Andrachek, et. al., January1996.10. WCAP-15293, Revision 2, Sequoyah Unit 1 Heatup and Cooldown Limit Curves for NormalOperation and PTLR Support Documentation, J.H. Ledger, July 2003.11. Westinghouse Letter to TVA, TVA-93 -105, Cold Overpressure Mitigation System Code Case andDelta-P Calculation, dated May 19, 1993.12. Calculation SQN-IC-01 4, Demonstrated Accuracy Calculation for Cold Overpressure ProtectionSystem.19 PRESSURE TEMPERATURE LIMITS REPORT13. ASME Code Case N-640, Alternative Reference Fracture Toughness for Development of P-TLimit Curves for Section XL, Division 1, dated February 26, 1999.14. WCAP-15984-P, Revision 01, Reactor Vessel Closure Head/Vessel Flange RequirementsEvaluation for Sequoyah Units 1 and 2, W. Bamford, et.al., April 2003.20 ENCLOSURE 3SEQUOYAH UNIT 2PRESSURE TEMPERATURE LIMITS REPORT, REVISION 6 Bi88 151.016"PRESSURE TEMPERATURE LIMITs REPORT801IContractrfo ny port of his re-.dligdealsad dinmsnsions. .*: 'Octobe~r 16,.2015. _* SOEP (Nt) BYW. J.Pierce__*Tennessee Valley AuthoritySequoyah unit 2Pressure Temperature Limits ReportRevision 6, september 2015PROJECT Secjuovah 'DisCIPLINE NCONTRACT " 4411 UNIT. 2DESC. RCS Pressure-Temperature L'mait-ReportDWG/DOC NO. PTLR-2SHEET ..- OF -REV. 06*DATE. 10/16/15 ECN/DCN FILE N2N-081"EDMS, WT CA-K PRESSURE TEMPERATURE LIMITS REPORTTable of ContentsLito T b e .List. .................of.............Tables........... .......................iList of Figures ............................................................................................. ........ v1.0 RCS Pressure Temperature Limits RePort (PTLR) ........ .... .................................... 12.0 Operating Limits ............................................................................. .............12.1 RCS Pressure/Temperature (PIT) Limits (TS 3.4.3)................................................. 13.0 Low Temperature Overpressure Protection System (TS 3.4.12) ...................... ..13.1 Pressurizer PORV Lift Setting Limits ........... .....................3.2. Arming Temperature ............'......24.0 Reactor Vessel Material Surveillance Program.....,................................................. 25.0 Supplemental Data Tables ............................................................................3..36.0 References ......................................... .. ... ..... ........ 17ii PRESSURE TEMPERATURE LIMITS REPORTList of TablesTable 2-1 Sequoyah Unit 2 Heatup Limits at 32 BEFPY(with Uncertainties for Instrumentation Errors of l0&deg;F and 60 psig)............................... 6Table 2-2 Sequoyah Unit 2 Cooldown Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 10&deg;F and 60 psig)...... ......................... 7Table 3.-1 Selected Setpoints, Sequoyah Unit 2........................................................... 8Table 4-1 Sequoyah Unit 2 Reactor Vessel Surveillance Capsule Withdrawal Schedule............. 10Table 5-1 Comparison of the Sequoyah Unit 2 Surveillance Material 30 ft-lb TransitionTemperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99,revision 2, Predictions................ .......................................................... 11Table 5-2 Calculation of Chemistry Factors using Sequoyah Unit 2 Surveillance Capsule Data....12Table 5-3 Reactor Vessel Beltline Material Unirradiated Toughness Properties forSequoyah Unit2 ................................................................................ 13Table 5-4 Peak Neutron Pluence Projections at Key Azimuthal Locations on the Reactor VesselClad/Base Metal Interface for Sequoyah Unit 2 (x 1019 n/cm2, EB> 1.0 MeV) ....i........ 14Table 5-5 Sequoyah Unit 2 Calculation of the ART Values for the 1/4T Location @ 32 EFPY.....15Table 5-6 Sequoyah Unit 2 Calculation of the ART Values for the 3/4T Location @ 32 EPPY.....15Table 5-7 Summary of the Limiting ART Values Used in the Generation of the Sequoyah Unit 2Heatup/Cooldown Curves.......................................................... :............ 16Table 5-8 RTPTs Calculations for Sequoyah Unit 2 Beltline Region Materials at 32 EFPY.......... 16iii PRESSURE TEMPERATURE LIMITS REPORTList of FiguresFigure 2-1 Sequoyah Unit 2 Reactor Coolant System Heatup Limitations(Heatup Rate of I 00&deg;F/hr) Applicable for the First 32 EFPY(wi/Margins for'Instrumentation Errors of 1 00F and 60 psig)................................ 4Figure 2-2 Sequoyah Unit 2 Reactor Coolant System Cooldown Limitations(Cooldown Rates up to 1 00&deg;F/hr) Applicable for the First 32 EFPY(w/ Margins for Instrumentation Errors of 10&deg;F and 60 psig) .................... .............5Figure 3-1 Sequoyah Unit 2 Selected LTOPS Setpoints .................................................. 9iv PRESSURE TEMPERATURE LIMITS REPORT1.0 RCS Pressure Temperature Limits Report (PTLR)This PTLR for Sequoyah Unit 2 has been prepared in accordance with the requirements of TechnicalSpecification (TS) 5.6.4. Revisions to the PTLR shall be provided to the NRC after issuance.This report affects TS.3.4.3, RCS Pressure/Temperature Limits (P/T) Limits and TS 3.4.12,.LowTemperature Overpressure Protection (LTOP) System.2.0 RCS Pressure and Temperature LimitsThe limits for TS 3.4.3 are presented in the subsections which follow and were developed using the NRCapproved methodologies specified in TS 5.6.4 with exception of ASME Code Case N-64011n1 (Use of K1o),WCAP- 159 84-Ptl21 (Elimination of the Flange Requirement), 1996 Version of Appendix G[41 and therevised fluencesETI. The operability requirements associated with LTOPS are specified in TS 3.4.12 andwere determined to adequately protect the RCS against brittle fracture in the event of an LTOP Transientin accordance with the methodology specified in TS 5.6.4.2.1 RCS Pressure/Temperature (P/T) Limits (TS 3.4.3)2.1.1 The minimum boltup temperature is 50&deg;F2.1.2 The RCS temperature rate-of-change limits are:a. A maximum heatup rate of 1 00&deg;F in any one hour period.b. A maximum cooldown rate of 100&deg;F in any one hour period.c. A maximum temperature change of less than or equal to. 1 0&deg;F in any one hour period duringinservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.2.1.3 The RCS P/T limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticalityare specified by Figures 2-1 and 2-2.3.0 Low Temperature Overpressure Protection System (TS 3.4.12)The lift setpoints for the pressurizer Power Operated Relief Valves (PORVs) are presented in thesubsections which follow. These lift setpoints have been developed using the NRC-approvedmethodologies specified in TS 5.6.4.I PRESSURE TEMPERATURE LIMITS REPORT*3.1 Pressurizer PORV Lift Setting LimitsThe pressurizer PORV lift setpoints are specified by Figure 3-1 and Table 3-1 (Ref. 10). The limits for theLTOPS setpoints are contained in the 32 EFPY steady-state curves (Table 2-2), which are beitlineconditions and are not compensated for pressure differences between the pressurizer transmitter and thereactor midplane/beltline or for instrument inaccuracies. The pressure difference between the pressurizertransmitter and the reactor vessel midplane/beltline with four reactor coolant pumps in operation is 68.3psi (Ref. 13).Note: These set-points include allowance for the pressure difference between the pressurizertransmitter and the reactor vessel midplane/beltline and the 50&deg;F thermal transport effect forheat injection transients.. A demonstrated accuracy calculation (Reference 14) has beenperformed to confirm that the setpoints will maintain the system pressure within theestablished limits when the pressure difference between the pressure transmhitter and reactormidplane and maximum temperature/pressure instrument uncertainties are applied to thesetpoints.3.2 Arming TemperatureThe LTOPS arning temperature is based upon the methodology defined in the Sequoyah Nuclear PlantUnit 2 Technical Specifications Administrative Controls Section 5.6.4. The arming temperature shall be <350&deg;F.4.0 Reactor Vessel Material Surveillance ProgramThe reactor vessel material irradiation surveillance specimens shall be removed and examined todetermine changes in material properties. The removal schedule is provided in Table 4-1. The results ofthese examinations shall be used to update Figures 2-1, 2-2 and 3-1.The pressure vessel steel surveillance program (WCAP-85 13[1]) is in compliance with Appendix H to 10CFR 50, "Reactor Vessel Material Surveillance Program Requirementst21."' The material test requirementsand the acceptance standard utilize the reference nil-ductility temperature RTNDT, which is determined inaccordance with ASTM E23 [3. The empirical relationship between RTNOT and the fracture toughness ofthe reactor vessel steel is developed in accordance with Code Case N-640 of Section XI of the ASMEBoiler and Pressure Vessel Code, Appendix G, "Fracture Toughness Criteria for Protection AgainstFailurer4k.' The surveillance capsule removal schedule meets the requirements of ASTM E185-82N5.Theremoval schedule is provided in Table 4-1.2 PRESSURE TEMPERATURE LIMITS REPORT5.0 Supplemental Data TablesTable 5-1 contains a comparison of measured surveillance material 30 ft-lb transition temperature shiftsand upper shelf energy decreases with Regulatory Guide 1.99, Revision 2[6], predictions.Table 5-2 shaows calculations of the surveillance material chemistry factors using surveillance capsuledata. Note that in the calculation of the surveillance weld chemistry factor, the ratio procedure fromRegulatory Guide 1.99, Revision 2 was followed. The ratio in question is equal to 0.93.Table 5-3 provides the required Sequoyah Unit 2 reactor vessel toughness data.Table 5-4 provides a summary of the fluence values used in the generation of the heatup and cooldownlimit curves and the PTS evaluation.Table 5-5 and 5-6 show the calculation of the 1/4T and 3/4T adjusted reference temperature at 32 EFPYfor each beltline material in the Sequoyah Unit 2 reactor vessel. The limiting beltline material was theIntermediate Shell Forging 05..Table 5-7 provides a summary of the adjusted reference temperature (ART) values of the Sequoyah Unit 2reactor vessel beltline materials at the l/4T and 3/4T locations for 32 EFPY.Table 5-8 provides RTP-Ts values for Sequoyah Unit 2 at 32 EFPY.3 PRESSURE .TEMPERATURE LIMITS REPORTMATERIAL PROPERTY BASISLIMITING MATERIAL! INTERMEDIATE SHELL FORGING 05LIMITING ART VALUES AT 32 EFPY: 1/4T, 142&deg;F3/4T, l15&deg;F2Rflfl-v Io)perlim Version:5.1 Run:5694 /ILeak, Test, Limit 22502000--[ Unacceptable1Oeain ....*Acceptable ...Operation1750 -..1500125U) 00Co0 25~ritical Limit*100 Deg. F/Hr750"500250MinumumBoltup Temp-- = 500F "Criticality Limit based on-~inservic e hydrostatic testtemperature (214&deg;F) for theservice period up to 32 EFPYUf.......I ..............! ....! ........ .........0 50 100 150 200 250 300 350 400 450 500 550Moderator Temperature (Deg. F)Figure 2-1 Sequoyah Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of1000F/hr) Applicable for the First 32 EFFY (w/ Margins for InstrumentationError of 10&deg;F and 60 psig) (Plotted Data provided on Table 2-1)
PRESSURE TEMPERATURE LIMITS REPORTMATERIAL PROPERTY BASISLIMITING MATERIAL: INTERMEDIATE SHELL FORGING 05.LIMITING ART VALUES AT 32 EFPY: 1/4T, 142&deg;F3/4T, 150F.2500 [Oeri Version:5.1 Run:5694]2000 tUnacceptable _____ --__ _________ AcceptableOperation OPeration -1750 ..... 'S 1250 * -.-o----__ -__ __ ____ '--"o Cooldown~~Rates i"= 1000 F/Hr .......................u steady-stateo-6o1750 60 oo250 -.... ...........M inim um I- ......:.. ........ .. ... ...4.. ... .........SBoltup Templi ..! ' i-500F !0 50 100 150 200 250 300 350 400 450 500 550Moderator Temperature (Deg. F)Figure 2-2 Sequoyahi~nit 2 Reactor Coolant System Cooldown Limitations (CooldownRates up to 100&deg;F/hr) Applicable for the First 32 EFPY (w/ Margins forInstrumentation Error of 100F and 60 psig) (Plotted Data provided on Table 2-2)2 PRESSURE TEMPERATURE LIMITS REPORTTable 2-1Sequoyah Unit 2 Heatup Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 10&deg;F and 60 psig)100 Heatup 1100 Critical Limit ILeak Test LimitT P T P JT P50 0 214 0 198 200050 591 214 607 214 248555 595 214 61460 601 214 62265 607 214 65770 614 214 65075 622 214 64780 630 214 64685 640 214 64890 646 214 *65395 646 214 661100 6.46 214 671105 646 214 680110 646 214 685115 646 214 701120 646 214 720125 648 214 743130 653 214 769135 661 214 798140 671 215 832145 685 220 869150 701 225 911155 720 230 959160. 743 235 1011165 769 240 1069170 798 245 1134175 832 250 1206180 869 255 1286185 911 260 1374190 959 265 1471195 1011 270 1579200 1069 275 1698205 1134 280 1829210 1206 285 1974215 1286 290 2134220 1374 295 2311225 1471230 1579235 1698240 1829245 1974250 2134255 23113 PRESSURE TEMPERATURE LIMITS REPORTTable 2-2Sequoyah Unit 2 Cooldown Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 1 0&deg;F and 60 psig)Steady State 1 20F 40F 60F I00OFT P T P T P T P T P5050556065707580859095100105110115120125130135140145*1501551601651761751801851901952002052102152202252300591595601607614622630640650661674688703720739.7607838098378689029409821028108011361199126813441429152216251739186520042158232850505560657075808590951001051101151201251301351401451501551601650552554558564572580589599610623636652668687707730755783814848885927.973102450505560657075808590951001051101151201251301351401451501551600503508514521529538548*55957158459961663465467670172975979383187291896850505560657075808590951001051101151201251301351401451501550.461466470478486496506518531546562580600622647674704738775816862913505055606570758085909510010511011512012513013514014515003663723803893994104234374534704905125365635936266637047498008564 PRESSURE TEMPERATURE LIMITS REPORTTable 3-1Selected Setpoints, Sequoyah Unit 2PORV#2 PORV#1Trcs (Deg.F) Setpoint (psig) Setpoint (psig)50 510 485100 580 555135 640 610174 745 682200 745 685*250 745 *685-278 745 685400 745 685450 2350 23505 PRESSURE TEMPERATURE LIMITS REPORTSequoyah Unit 2 LTOPS Selected Setpoints-/! [4-2500 --....J Lo- I I !0 01015 0 5 303040 5 0Reacto Coln Iyte enpraue(FA F--*-- FORV#2-FinaI --U-- FORV#1Figure 3-1 Sequoyah Unit 2 Selected LTOPS Setpoints (Plotted Data pro vided on Table 3-1)6 PRESSURE TEMPERATURE LIMITS REPORTTable 4-1Sequoyah Unit 2 Reactor Vessel Surveillance Capsule Withdrawal Schedule(a) Updated in Capsule Y dosimetry analysis (WCAP-15320t71).(b) Effective Full Power Years (EFPY) from plant startup.(c) Plant specific evaluation.(d) This fluence is not less than once or greater than twi ce the peak end of license (32 EFPY) fluence(e) Capsules 5, V, W and Z will reach a fluence of 2.71 x 1019 (EB> 1.0 MeV), the 48 EFPY peakvessel fluence at approximately 44 EFPY.,Administrative Note -The surveillance capsule withdrawal schedule in Table 4-1 is based on thesurveillance program for the original 40 year service life. Relocation of select standby, capsules toincrease the fluence lead factor in anticipation of the updated surveillance program for the 60 year licenserenewal service life is described in a TVA letter to NRC dated January 10, 2013 (ML13032A25 1). .Regulatory approval for the anticipatory standby capsule relocation has been granted by a NRC letter toTVA dated September 27, 2013 (ML13240A320). A complete update of the reactor vessel surveillanceprogram for the 60 year license renewal service life will be documented by a subsequent revision to thePTLR prior to entry into the license renewal extended operating period.7 PRESSURE TEMPERATURE LIMITS REPORTTable 5-1Comparison of the Sequoyah Unit 2 Surveillance. Material 30 ft-lb Transition Temperature Shifts andUpper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions30 ft-lb Transition Upper Shelf EnergyTemperature Shift DecreaseMaterial Capsule Fluence ' Predicted Measured Predicted Measured(x 1019 n/cm2) (OF)(a) (OF)(b) (%,/)(al) (%)(c)Intermediate Shell T 0.261 60.33. 63.65 17 12Forging 05 U 0.692 85.22 79.31 21 16(Tangential)(et285/X 1.22 100.23 85.7 23 8981057) Y 2.14 114.67 134.12 26 22Intermediate Shell T 0.261 60.33 48.73 17 7Forging 05 U 0.692 85.22 66.06 21 9(Axial)(et285/X 1.22 100.23 110.04 23 2981057) Y 2.14 .114.67 89.21 26 22Weld Metal T 0.261 43.12 74.56 20 2(Heat # 4278)(d) U 0.692 60.91 130.38 25 6X 1.22 71.63 44.22 29 35Y 2.14 81.96 86.91 33 "3HAZ Metal T 0.261 --24.58 -,- 2U 0.692 --64.03 --14X 1.22 --28.29 --19Y 2.14 --50.32 --39Notes:(a) Based on Regulatory Guide 1.99, Revision 2, methodology using the mean weight percent values ofcopper 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) Surveillance Weld was fabricated from weld wire type SMIT 89, Heat # 4278, Flux type SMIT 89,Lot # 1211.8 PRESSURE TEMPERATURE LIMITS REPORTTable 5-2Calculation of Chemistry Factors using Sequoyah Unit 2 Surveillance Capsule DataMaterial Capsule Capsule f(a) FF(b) ARTNDT(C) FF*ARTNDT FF2Intermediate Shell T 2.61E+18 0.635 63.7 40.45 0.403Forging 05 U 6.92E+18 0.897 79.3 71.13 0.805(Tangential) X 1.22E+19 1.055 85.7 90.41 1.113(Heat #288757 / Y 2.14E+19 1.207 134.1 161.86 1.457981057)Intermediate Shell T 2.61E+18 0.635 48.7 30.92 0.403Forging 05 U 6.92E+18 0.897 66.1 -59.29 0.805(Axial) X 1.22E+19 1.055 110.0 116.05 1.113(Heat #288757 / Y 2.14E+19 1.207 89.2 107.66 1.457981057) ______ ______SUM: 677.77&deg;F 7.556CFo5 = X(FF
* RTNDT) + X.( FF2) = (677.77) +(7.556) = 89.70FSurveillance Weld T 2.6 1E+18 0.635 69.4 (74.6) 44.07 0.403Material(d) *U 6.92E+18 0.897 121.3 (130.4) 108.81 0.805(Heat # 4278)(e) X 1.22E+19 1.055 41.1 (44.2) 43.36 1.113Y 2.14E+19 1.207 80.8 (86.9) 97.53 1.457SUM: 293 .77&deg;F 3.778CF Surv. weld =
* RTNDT) + X( FF2) = (293.77&deg;F) +(3.778) = 77.8"FNotes:(a) f = Calculated fluence from Capsule Y dosimetry analysis results [7], (n/cma2, E > 1.0 MeV).(b) FF = fluence factor = f02-.~o 3(c) AIRTNDT values are the measured 30 ft-lb shift values taken from App. B of Ref. 7, rounded to onedecimal point.(d) The surveillance weld metal ARTNDT Values have been adjusted by a ratio factor of 0.93.(e) Surveillance Weld was fabricated from weld wire type SMIT 89, Heat # 4278, Flux type SMIT 89, Lot #1211.
PRESSURE TEMPERATURE LIMITS REPORTTable 5-3Reactor Vessel Beltline Material Unirradiated TouglmessProperties for Sequoyah Unit 2Material Description Cu (%) Ni (%) Initial RTNDT(a)Intermediate Shell Forging 05(Heat #288757 /981057)013.710FLower Shell Forging 040.14 0.76 -22 &deg;F(Heat # 990469 / 293323)Intermediate to Lower Shell ForgingCircumferential Weld Seam W05(b) 0.12 0.11 -40F(Heat # 4278)Surveillance Weld(b) 0.13 0.11Notes:(a) The Initial RTNDT values are measured values(b) Circumferential Weld Seam was fabricated with weld wire type SMIT 89, Heat # 4278, Flux type SMIT89, Lot # 1211 and is representative of the intermediate to lower shell circumferential weld.10 PRESSURE TEMPERATURE LIMITS REPORTTable 5-4Peak Neutron Fluence Projections at Key Azimuthal Locations on the Reactor Vessel Clad/Base MetalInterface for Sequoyah Unit 2 (x 1019 n/cm2, E > 1.0 MeV)Azimuthal LocationEFPY 0&deg; 150 3 00 45010.54 0.211 0.336 0.426 0.63720 0.38 0.60 0.773 1.1632 0.593 0.934 1.21 1.8248 0.878 1.38 1.80 2.7111 PRESSURE TEMPERATURE LIMITS REPORTTable 5-5Sequoyah Unit 2 Calculation of the ART Values for the 1/4T Location @ 32 EFPY(a)Material RG 1.99 cF FF IRTNDT(b) ARTNDTo(C) Margin(d) ART(e)R2 Method (0F) _____ (0F) (0F) (0F) (0F)Position 1.1 95 1.027 10 97.6 34 142Intermediate Shell Forging 05Position 2.1 89.7 1.027 10 92.1 34 136Lower Shell Forging 04 Position 1.1 104 1.027 -22 106.8 34(0 119Intermediate to Lower Shell " Position 1.1 .63 '1.027 -4 64.7 56 117Circumferential Weld Seam Position 2.1 77.8 " 1.027 .-4 79.9 56(0 132Notes:(a) Neutron fluence value used for all materials is the highest value from Table 5-4 for 32 EFPY.(b) Initial RTNDoT values are measured values.(c) ARTrmT CF *FF(d) M =2 *(a'i2 +I 0'A2)1/2(e) ART =Initial RTNDT +- ARTNDTr + Margin (0F)(f) Data deemed not-credible (See Reference 7a), thus the full ca will be used to determine margin.Table 5-6Sequoyah Unit 2 Calculation of the ART Values for the 3/4T Location @ 32 EFPY(a)Material RG 1.99 CF FF IRTNDT(b) Margin(d) ART(e)R2 Method (0F) (0F) (0F) (0F) (0F)Position 1.1 95 0.745 10 70.8 34
* 115Intermediate Shell Forging 05Position 2.1 89.7 0.745 10 66.8 34 .111Lower Shell Forging 04 Position 1.1 104 0.745 -22 77.5 34(0 90intermediate to Lower Shell Position 1.1 63 0.745 .-4 46.9 56 99Circumferential Weld Seam Position 2.1 77.8 0.745 -4 58.0 56(0 110Notes: .(a) Neutron fluence value used for all materials is the highest value from Table 5-4 for 32 EFPY.(b) Initial RTrNDT values are measured values.(c) =CF *FF(d) M = 2 *(0i2 + raA)2(e) ART =Initial RTrNDv + ARTNDT + Margin (0F)(f) Data deemed not-credible (See Reference 7a), thus the full GA will be used to determine margin.12 PRESSURE TEMPERATURE LIMITS REPORTTable 5-7Summary of the Sequoyah Unit 2 Reactor Vessel Beitline Material ART ValuesMaterial RG 1.99 R2 1/4 ART 3/4 ARTMethod (0F) (0F)Position 1.1 142 115Intermediate Shell Forging 05Position 2.1 136 111Lower Shell Forging 04 Position 1.1 119 90Intermediate to Lower Shell Position 1.1 117 99Circumferential Weld Seam Position 2.1I 132 110Table 5-8RTPTs Calculations for Sequoyah Unit 2 BeltlineRegion Materials at 32 EFPY(a)Material Fluence FF CF ARTpTs~b) Margin RTNDT(U-)(C) RTpTs(d)(x 10 19 n/cm2, (0F) (0F) (0F) (0F) (0F)E>l.0 MeV)Intermediate Shell Forging 05 1.82 1.164 95 110.6 34 10 155Intermediate Shell Forging 05 1.82 1.164 89.7 104.4 34 10 148(Using S/C Data) _______Lower Shell Forging 04 1.82 .1.164 104 121.1 34(e) -22 133Circumferential Weld Metal 1.82 1.164 63 73.3 56 -4 125Circumferential Weld Metal 1.82 1.164 77.8 90.6 56(e) -4 143(Using S/C Data)____Notes:(a) Neutron fluence value used for all materials is the highest value from Table 5-4 for 32 EFPY.(b) ARTpTs= CF
* FF(c) Initial RTNDT values are measured values(d) RTpTs = RTNDT(U) + ARTpTs + Margin (0F)(e) Data deemed not-credible (See Reference 7a), thus the full ra.. will be used to determine margin.13 PRESSURE TEMPERATURE LIMITS REPORT6.0 References1. WCAP-8513, Tennessee Valley Authority Sequoyah Unit No. 2 Reactor Vessel RadiationSurveillance Program, J. A. Davidson, et. al., November 1975.2. Code of Federal Regulations, 10OCFR50, Appendix H, Reactor Vessel Material SurveillanceProgr~am Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C.3. ASTM E23 Standard Test Method Notched Bar" Impact Testing of Metallic Materials, in ASTMStandards, American Society for Testing and Materials, Philadelphia, PA.4. Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G, Fracture ToughnessCriteria for Protection Against Failure5. ASTM E185-82, Annual Book of ASTM Standards, Section 12, Volume 12.02, Standard Practicefor Conducting Surveillance Tests for Light- Water Cooled Nuclear Power Reactor Vessels.6. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S.Nuclear Regulatory Commission, May 1988.7a. WCAP- 15320, Analysis of Capsule Y fi'omn the Tennessee Valley Authority Sequoyah Unit 2Reactor Vessel Radiation Surveillance Program, T.J. Laubham, et. al., November 1999.7b. WCAP-*10509, Analysis of Capsule T fiom the. Tennessee Valley Authority Sequoyah Unit 2Reactor Vessel Radiation Surveillance Program, R. S. Boggs, et al, April 1984.7c. Southwest Research Institute Nondestructive Evaluation Science and Technology Division,Reactor Vessel Material Surveillance Program and Technology Division, Reactor Vessel Material"Surveillance Pro gram for Sequoyah Unit 2." Analysis of Capsule U, Final Report SwRI Project*17-8851 TVA Contra~ct 85PJH-964430, January !1990.7d. WCAP- 13545, Analysis of Capsule X from the Tennessee Valley Authority Sequoyah Unit 2Reactor Vessel Radiation Surveillance Program, M. A. Ramirez, S. L. Anderson, A. Madeyski,November 1992.8. CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.1, developed by ATIConsulting, March 1999.9. WCAP- 14040-NP-A, Revision 2, Methodology Used to Develop Cold Overpressure MitigatingSystem Setpoints and RCS Heatup and Cooldown Limit Curves, J.D. Andrachek, et. al., January1996.10. WCAP-15321, Revision 2, Sequoyah Unit 2 Heatup and Cooldown Limit Curves for NormalOperation and PlTLR Support Documentation, J.H. Ledger, et.al., July 2003.11. ASME Code Case N-640, Alternative Reference Fracture Toughness for Development of P-T*Limit Curves for Section XL, Division 1, dated February 26, 1999.14 PRESSURE TEMPERATURE LIMITS REPORT12. WCAP-15984-P, Revision 01, Reactor Vessel Closure Head/Vessel Flange RequiremnentsEvaluation For Sequoyah Units 1 and 2, W. Bamford, et.al., April 2003.13. Westinghouse Letter to TVA, TVA-93-105, Cold Overpressure Mitigation System Code Case andDelta-P Calculation, dated May 19, 1993.14. Calculation SQN-IC-0 14, Demonstrated Accuracy Calculation for Cold Overpressure ProtectionSystem.15}}

Revision as of 19:26, 2 June 2018

Sequoyah, Units 1 and 2, Transmittal of Pressure Temperature Limits Report, Revision 5 and 6
ML15322A055
Person / Time
Site: Sequoyah  Tennessee Valley Authority icon.png
Issue date: 11/13/2015
From: Carlin J T
Tennessee Valley Authority
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML15322A055 (51)
v • d • e


Text

Tennessee Valley Authority, Post Office Box 2000, Soddy Daisy, Tennessee 37384-2000November 13, 201510 CFR 50.4ATT-N: Document Control DeskU.S. Nuclear Regulatory CommissionWashington, D.C. 20555-0001Sequoyah Nuclear Plant, Units I and 2Renewed Facility Operating License Nos. DPR-77 and DPR-79NRC Docket Nos. 50-327 and 50-328

Subject:

Sequoyah Unit I Pressure Temperature Limits Report,Revision 5, and Sequoyah Unit 2 Pressure Temperature LimitsReport, Revision 6

References:

1. Letter from NRC to TVA, "Sequoyah Nuclear Plant, Units I and 2-Issuance of Amendments for the Conversion to the ImprovedTechnical Specifications with Beyond Scope Issues (TAC Nos.MF3128 and MF3129)," dated September 30, 20015(MLI15238B460)In accordance with Sequoyah Nuclear Plant (SQN) Units 1 and 2 TechnicalSpecifications (Tss) 5.6.4.c, enclosed is the Unit I Pressure Temperature LimitsReport (PTLR), Revision 5, and Unit 2 PTLR, Revision 6. In accordance with TSs5.6.4.c, the PTLRs are required to be provided to the Nuclear RegulatoryCommission (NRC) within 30 days after any revision. Sequoyah Units I and 2 wereissued license amendment Nos. 334 and 327, respectively for improved standardTSs (Reference 1). These license amendments resulted in the enclosed revisions toeach of the PTLRs. The revisions also include other editorial clarifications andadministrative changes identified during the revision process as described inEnclosure 1. The revised PTLRs became effective on October 16, 2016.There are no new regulatory commitments in this letter. If you have any questions,please contact Jonathan Johnson, SQN Site Licensing Manager at (423) 843-8129.

U.S. Nuclear Regulatory CommissionPage 2November 13, 2015Sequo Nuclear PlantEnclosures1.2.3.Units 1 and 2 Pressure Temperature Limits Report ChangesSequoyah Unit 1 Pressure Temperature Limits Report, Revision 5Sequoyah Unit 2 Pressure Temperature Limits Report, Revision 6ZTK: DVGEnclosurescc (Enclosures):NRC Regional Administrator -Region IINRC Senior Resident Inspector -SQN ENCLOSURE1ISEQUOYAH UNITS 1 AND 2PRESSURE TEMPERATURE LIMITS REPORT CHANGESThe following describes the editorial clarifications and administrative changes madeto each Units' Pressure Temperature Limits Report (PTLR).1. Conflict between the figure index referencing Cold Overpressure MitigationSystem (COMS) and the figure title referencing Low TemperatureOverpressure Protection System (LTOPS) was Corrected for consistency withthe system terminology in Section 3.4.12 of the Technical Specifications.2. Section 1, "RCS Pressure Temperature Limits Report (PTLR)," was revisedto clarify which Limiting Condition for Operations are affected by the PTLR.3. Section 3.1, "Pressurizer PORV Lift Setting Limits," was revised forconsistency with the analysis of record contained in Topical Report No.WCAP-1 5293, Revision 2 "Sequoyah Unit 1 Heatup and Cooldown LimitCurves for Normal Operation and PTLR Support Documentation," andTopical Report No. WCAP-15321 Revision 2 "Sequoyah Unit 2 Heatup andCooldown Limit Curves for Normal Operation and PTLR SupportDocumentation".4. Section 5.0, "Supplemental Data Tables," added the word "Forging" to thesentence in refer to the limiting beltline material identified in Tables 5-5 and5-6.5 A note was added to Table 4-1, "Sequoyah Unit 1 Reactor VesselSurveillance Capsule Withdrawal Schedule," to describe current and futureactivity with the surveillance capsule withdrawal schedule.6. Tables 5-5 and 5-6 were revised with a different annotation conventionconsistent with the other tables in the PTLR.

ENCLOSURE 2SEQUOYAH UNIT 1PRESSURE TEMPERATURE LIMITS REPORT, REVISION 5 B88 15 t 16800MAPRESSURE TEMPERATURE LIMITS REPORT....APPROVEDThis qp~provaIl does. not reliev, theContractor from any port o9 his re-,*ponsibility Jar, the correctness of.* design. debugailnd dimlnsio~ns.YLetter lo.Nl107l6.*oot,: ,October 16, 2015. -.SQEP BY: W. JX PierceTennessee Valley AuthoritySequoyah unit 1Pressure Temperature Limits ReportRevision 5, September 2015PROJECT Sequovah .DISCIPLINE N"CONTRACT .4411 1JNT' 1DESC. RCS Pressure-Temperature Limit Report*DWG/DOC NO. PTLR-1SHEET -OF -. "REV. 05*DATE 10/16/15 ECN/DCN -. FILE N2N-081EDMS, WT CA-K PRESSURE TEMPERATURE LIMITS REPORTTable of ContentsList of Tables .................................................................................................... ivList of Figures......................................................................................................v1.0 RCS Pressure Temperature Limits Report (PTLR) ......................................... ....... 12.0 Operating Limits.......................................................................................... 12.1 RCS Pressure/Temperature (PIT) Limits (TS 3.4.3) ............................................... 13.0 Low Temperature Overpressure Protection System (TS 3.4.12)...... ............................ 13.1 Pressurizer PORV Lift Setting Limits .............................................................. 23.2 Arming Temperature................................................................................ 24.0 Reactor Vessel Material Surveillance Program...................................................... 25.0 Supplemental Data Tables............................................................................... 36.0 References................................................................................................ 19ii PRESSURE TEMPERATURE LIMITS REPORTList of TablesTable 2-1 Sequoyah Unit 1 Heatup Limits at 32 EPPY(with Uncertainties for Instrumentation Errors of 10°F and 60 psig)......... ...................... 6Table 2-2 Sequoyah Unit 1 Cooldown Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 1 0°F and 60 psig)................................STable 3-1 Selected Setpoints, Sequoyah Unit 1.......................................................... 10Table 4-1 Sequoyah Unit I Reactor Vessel Surveillance Capsule Withdrawal Schedule ............ 12Table 5-1 Comparison of the Sequoyah Unit 1 Surveillance Material 30 ft-lb TransitionTemperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99,revision 2, Predictions .............................'............................................. 13Table 5-2 Calculation of Chemistry Factors using Sequoyah Unit 1 Surveillance Capsule Data. .... 14Table 5-3 Reactor Vessel Beltline Material Unirradiated Toughness Properties forSequoyah Unit 1 ............................................................................ ..... 15Table 5-4 Peak Neutron Fluence Projections at Key Azimuthal Locations on the Reactor VesselClad/Base Metal Interface for Sequoyah Unit 1 (x 10'9 n/cm2, B > 1.0 MeV) ............ 16Table 5-5 Sequoyah Unit 1 Calculation of the ART Values for the 1 /4T Location @ 32 EFPY.....17Table 5-6 Sequoyah Unit 1 Calculation of the ART Values for the 3/4T Location @ 32 EFPY ......17Table 5-7 Summary of the Limiting ART Values Used in the Generation of the Sequoyah Unit 1Heatup/Cooldo~vn Curves. .................................................................... 18Table 5-8 RTm-s Calculatiohis for Sequoyah Unit 1 Beitline Region Materials at 32 EFPY.......... 18iii PRESSURE TEMPERATURE LIMITS REPORTList of FiguresFigure 2-1 Sequoyah Unit 1 Reactor Coolant System Heatup Limitations(Heatup Rate of 1 00°F/hr) Applicable for the First 32 EFPY(w/ Margins for Instrumentation Errors of 1 0°F and 60 psig)................................ 4Figure 2-2 Sequoyah Unit 1 Reactor Coolant System Cooldown Limitations(Cooldown Rates up to 100°F/hr) Applicable for the First 32 EFPY(w/ Margins for Instrumentation Errors of 1 0°F and 60 psig)................................ 5Figure 3-1 Sequoyah Unit 1 Selected LTOPS Setpoints ................................................. 11iv PRESSURE TEMPERATURE LIMITS REPORT1.0 RCS Pressure Temperature Limits Report (PTLR)This PTLR for Sequoyah Unit 1 has been prepared in accordance with the requirements of TecirnicalSpecification (TS) 5.6.4. Revisions to the PTLR shall be provided to the NRC after issuance.This report affects TS 3.4.3, RCS Pressure/Temperature Limits (PiT) Limits and TS 3.4.12, LowTemperature Over Pressure Protection (LTOP) System.2.0 RCS Pressure and Temperature LimitsThe limits for TS 3.4.3 are presented in the subsections which follow and were developed using the NRCapproved methodologies specified in TS 5.6.4 with exception of ASME Code Case N-640[13] (Use of Kit),WCAP- 15984-P['14 (Elimination of the Flange Requirement), 1996 Version of Appendix Gm and the-revised fluencesE7T. The operability requirements associated with LTOPS are specified in TS 3.4.12 andwere determined to adequately protect the RCS against brittle fracture in the event of an LTOP Transientin accordance with the methodology specified in TS 5.6.4.2.1 RCS Pressure/Temperature (PIT) Limits (TS 3.4.3)2.1.1 The minimum boltup temperature is 500F2.1.2 The RCS temperature rate-of-change limits are:a. A maximum heatup rate of 1 00°F in any one hour period.* b. A maximum cooldown rate of l00OF in any one hour period.c. A maximum temperature change of less than or equal to I10°F in any one hour period duringinservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.2.1.3 The RCS P/T limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticalityare specified by Figures 2-1 and 2-2.3.0 Low Temperature Overpressure Protection System (TS 3.4.12)The lift setpoints for the pressurizer Power Operated Relief Valves (P.ORVs) are presented in thesubsections which follow. These lift setpoinlts have been developed using the NRC-approvedmethodologies specified in TS 5.6.4.1 PRESSURE TEMPERATURE LIMITS REPORT3.1 Pressurizer PORV Lift Setting LimitsThe pressurizer PORV lift setpoints are specified by Figure 3-1 and Table 3-1 (Ref. 10). The limits for theLTOPS setpoints are contained in the 32 EFPY steady-state curves (Table 2-2), which are beitlineconditions and are not compensated for pressure differences between the pressurizer transmitter and thereactor midplane/beltline or for instrument inaccuracies. The pressure difference between the pressurizertransmitter and the reactor vessel midplane/beltline with four reactor coolant pumps in operation is 68.3psi (Ref. 11).Note: These setpoints include allowance for the pressure difference between the pressurizertransmitter and the reactor vessel midplane/beltline and the SOTF thermal transport effect forheat injection transients. A demonstrated accuracy calculation (Reference 12) has beenperformed to confirn that the setpoints will maintain the system pressure within theestablished limits when the pressure difference between the pressure transmitter and reactormidplane and maximum temperature/pressure instrument uncertainties are applied to thesetpoints.3.2 Anning TemperatureThe LTOPS arming temperature is based upon the methodology defined in the Sequoyah Nuclear PlantUnit 1 Technical Specifications Administrative Controls Section 5.6.4. The arming temperature shall be _<350°F.4.0 Reactor Vessel Material Surveillance ProgramThe reactor vessel material irradiation surveillance specimens shall be removed and examined todetermine changes in material properties. The removal schedule is provided in Table 4-1. The results ofthese examinations shall be used to update Figures 2-1, 2-2 and 3-1.The pressure vessel steel surveillance program (WCAP-8233 r1l) is in compliance with Appendix H to 10CFR 50, "Reactor Vessel Material Surveillance Program Requirementsr21."' The material test requirementsand the acceptance standard utilize the reference nil-ductility temperature which is determined inaccordance with ASTM E23 [3]. The empirical relationship between RTNDr and the fracture toughness ofthe reactor vessel steel is developed in accordance with Code Case N-640 of Section XI of the ASMEBoiler and Pressure Vessel Code, Appendix 0, "Fracture Toughness Criteria for Protection AgainstFailureE41. The surveillance capsule removal schedule meets the requirements of ASTM E185-82N5.Theremoval schedule is provided in Table 4-1.2 PRESSURE TEMPERATURE LIMITS REPORT5.0 Supplemental Data TablesTable 5-1 contains a comparison of measured surveillance material 30 ft-lb transition temperature shiftsand upper shelf energy decreases with Regulatory Guide 1.99, Revision 2[6], predictions.Table 5-2 shows calculations of the surveillance material chemistry factors using surveillance capsuledata. Note that in the calcuilation of the surveillance weld chemistry factor, the ratio procedure fromRegulatory Guide 1.99, Revision 2 was followed. The ratio in question is equal to 0.90.Table 5-3 provides the required Sequoyah Unit 1 reactor vessel toughness data.Table 5-4 provides a summary of the fluence values used in the generation of the heatup and cooldownlimit curves and the PTS evaluation.Table 5-5 and 5-6 show the calculation of the 1/4T and 3/4T adjusted reference temperature at 32 EFPYfor each beitline material in the Sequoyah Unit 1 reactor vessel. The limiting beitline material was theLower Shell Forging 04.Table 5-7 provides a summary of the adjusted reference temperature (ART) Values of the Sequoyah Unit 1reactor vessel beltline materials at the 1/4T and 3/4T locations for 32 EFPY.Table 5-8Sprovides RTPTs values for Sequoyah Unit 1 at 32 EFPY.3 PRESSURE TEMPERATURE LIMITS REPORTMATERIAL PROPERTY BASISLTMITING MATERIAL: LOWER SHELL FORGING 04LIMITING ART VALUES AT 32 EFPY: 1/4T, 216°F3/4T, 186°F2500 --_o- erlim Version.5.1 Run.15680 [i /-2250i Leak Test Limit /-; -2000 .......Unacceptable ....... ... -- --"Acceptable ...,Operation, Operation__:Heatup Rate Critical Limit100o Deg. F/HrI 100 Deg. F/ HrI-- "~2=i(n 'i.120 ,.. .. ....S 1000 --- _ __- ___ __ _750 -- -- -__ _ _ -__ ___-__ __ __ __ --- __/ Criticality Limit based on Iinservice hydrostatic test500 -temnperatuire (2880F) for the __service period up to 32 EFPY(IMinimum !250 ......-~ Boltup .. ... -; -___ __) _STemp = 50*F0 50 100 150 200 250 300 350 400 450 500 550Moderator Temperature (Deg. F)Figure 2-1 Sequoyah Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of100°F/hr)-Applicable for the First 32 EFPY (w/Margins for Instrumentation Errorof 10°F and 60 psig) (Plotted Data provided on Table 2-1)4 PRESSURE TEMPERATURE LIMITS REPORTMATERIAL PROPERTY BASISLIMITING MATERIAL: LOWER SHELL FORGING 04LIMITING ART VALUES AT 32 EFPY: 114T, 216°F3/4T, 186°F2500225020001750150021250o 10007505002500 50 100 150 200 250300 350 400 450 500 550Moderator Temperature (Deg. F)Figure 2-2 Sequoyah Unit 1 Reactor Coolant System Cooldown Limitations (CooldownRates up to 100°F/hr) Applicable for the First 32 EFPY (w/Margins forInstrumentation Error of 10°F and 60 psig) (Plotted Data provided on Table 2-2)5 PRESSURE TEMPERATURE LIMITS REPORTTable 2-1Sequoyah Unit 1 Heatup Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 10°F and 60 psig)100 Heatup 1100 Critica] Limit ILeak Test LimitT P J T Pj T P50 0 288 0 272 200050 477 *288 477 288 248555 477 288 47760 477 288 47765 477 288 47770 477 288 47875 477 288 47880 477 288 48085 477 288 48190 477 ,288 48395 477. 288 485100 477 288 487105 477 288 490110 477 288 493115 477 288 497120 477 288 500125 477 288 505"130 477 288 508135 477 288 515140 477 288 517145 .477 288 527150 477 288 528155 478 288 541160 480 288 541165 483 288 555170 487 288 557175 493 288 571180 500 288 575185 508 288 589190 517 288 609195 528 288 631200 541 288 .656205 555 ,288 684210 571 288 714215 589 .288 748220 609 290 786225 631 295 8286 PRESSURE TEMPERATURE LIMITS REPORTTable 2-1 -(Continued)Sequoyah Unit 1 Heatup Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 1 0°F and 60 psig)100 Heatup 100 Critical LimitT P T P230 656 300 874235 684 305 925240 714 310 981245 748 315 1044250 786 320 1112255 828 325 1188260 874 330 1272265 925 335 1364270 981 340 1466275 1044 345 1578280 1112 350 1702285 1188 355 1838290 1272 360 1988295 1364 365 2154300 1466 370 2337305 1578310 1702315 1838320 1988325 2154330 23377 PRESSURE TEMPERATURE LIMITS REPORTTable 2-2Sequoyah Unit 1 Cooldown Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 10°F and 60 psig)Steady State [20F )40F P1 60F 100F50505560657075808590951001051101151201251301351401451501551601651701751801851901952002052102150552553555556558560561564566569571575578582586591596602608616623632642652664677691707724743764788814843505055606570758085,909510010511011512012513013514014515015516016517017518018519019520020521021505035055075095105125145165185215245275315355405455505565635715795885996106236376526696887097337597878195050556065707580859095100105110115120125130135140145150155160165170175180185190195200205210215045745845946046246446546847047347647948348749249750351051752553454455656858259761463365467770273176279750505560657075808590951001051101151201251301351401451501551601651701751801851901952002052102150408409410411412414416418420423426430434.4384434494564634714794895005125265415585775976206466747057407795050556065707580859095100105110115120125130'1351401451501551601651701751801851901952002052102150305306307308309311.313315318.321325329333338344351358367376"387399412427443461482505530558590624663706754______________________________________ +/- _______________________________________ _________________________________________ I8 PRESSURE TEMPERATURE LIMITS REPORTTable 2-2 -(Continued)Sequoy~a Unit 1 Cooldown Limits at 32 EFPY(without Uncertainties for Instrumentation Errors)Steady State 20F 40F 60F 100F.T P T P T PT P T P220 874 220 853 220 836 220 821 220 806225 909 225 892 225 878 225 869 2-25 865230 948 230 935 230 925 230 921235 991 235 982 235 978240 1038 240 1034245 1090250 1148255 1212260 1283265 1360270 1447275 1542280 1647285 1763290 1892295 2034300 2191305 23649 PRESSURE TEMPERATURE LIMITS REPORT*Table 3-1Selected Setpoints, Sequoyah Unit 1Trs Dg.) PORV#2 PORV#1Trcs(De.F)Setpoint (psig) Setpoint (psig)50 490 465100 500 475135 540 510175 575 540200 610 570250 745 685280 745 685405 745 685450 2350 235010 PRESSURE TEMPERATURE LIMITS REPORTSequoyah Unit I LTOPS Selected Setpoints250020002i1500100005000 50 100 150 200 250 300 350 400 450 500Reactor Coolant System Temnperature (°F)--4-- FOFRV#2 Setpoint --=- PORV#1 SetpointFigure 3-1 Sequoyah Unit 1 Selected LTOPS Setpoints (Plotted Data provided on Table 3-1)11 PRESSURE TEMPERATURE LIMITS REPORTTable 4-1Seqtuoyah Unit 1 Reactor Vessel Surveillance Capsule Withdrawal Schedule* Removal Time FluenceCapsule Location Lead Factorca) (EFPY) (b) (n/cm2,E>l.0 1T 400 3.39 1.03 2.61 x 10O8 (c)U 1400 3.47 3.00 7.96 x 1018 (c)X 2200 3.47 5.27 1.32 x 1019 (c)Y 320° 3.43 10.03 2.19 x 10'9 (c,d)5 40 1.08 Standby (d,e)V 176° 1.08 Standby (d,e)W 184° 1.08 Standby (d,e)Z 3560 1.08 Standby (d,e)Notes:(a)(b)(c).(d)(e)Updated in Capsule Y dosimetry analysis Effective Full Power Years (EFPY) from plant startup.Plant specific evaluation.This fluence is not less than once or greater than twice the peak end of license (32 EFPY) fluenceCapsules 5, V, W and Z will reach a fluence of 2.74 x 1019 (E > 1.0 Mev), the 48 EFPY peakvessel fluence at approximately 44 EFPY, respectively.Administrative Note -The surveillance capsule withdrawal schedule in Table 4-1 is based on thesurveillance program for the original 40 year service life. Relocation of select standby capsules toincrease the fluence lead factor in anticipation of the updated surveillance program for the 60 year licenserenewal service life is described in a TVA letter to NRC dated May 14, 2015 (ML15!34A377).Regulatory approval for the anticipatorY standby capsule relocation has been granted by a NRC letter toTVA dated September 4, 2015 (ML 15244B222). A complete update of the reactor vessel surveillanceprogram for the 60 year license renewal service life will be documented by a subsequent revision to thePTLR prior to entry into the license renewal extended operating period.12 PRESSURE TEMPERATURE LIMITS REPORTTable 5-1Comparison of the Sequoyah Unit 1 Surveillance Material 3 0 ft-lb Transition Temperature Shifts andUpper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions30 ft-lb Transition Upper Shelf EnergyTemperature Shift DecreaseMaterial Capsule Fluence Predicted Measured Predicted Measured____________ ______ (X 1019 n/cm2). (OF)(a) (OF)(b) (%)(a) .(%)(C)Lower Shell T 0.261 .59.85 67.52 16 16Forging 04 U 0.796 89.3 109.7 20.5 21(Tneta)X 1.32 102.6 145.12 23 8(Heat # 980919 /281587) Y 2.19 114.95 129.87 26.5 23Lower Shell T 0.261 59.85 50.59 16 0Forging 04 U 0.796 89.3 67.59 20.5 19(xa)X 1.32 102.6 103.34 23 22(Heat # 980919 / _____281587) Y 2.19 114.95 133.35 26.5 19Weld Metal T 0.261 111.13 127.79 .35 30(Heat # 25295)(d) U 0.796 165.82 144.92 42 26X 1.32 190.51 159.02 .45 .21Y 2.19 213.44 163.8 48 28IiAZ Metal T 0.261 --45.48 --20U 0.796 --. 78.94 --26X 1.32 --95.89 --3Y 2.19 --73.3 --10Notes:(a) Based on Regulatory Guide 1.99, Revision 2, methodology using the mean weight percent values ofcopper and nickel of the surveillance material.(b) .Calculated using measured Charpy data plotted using CVGRAPH, Version 4.1 [8(c) Values are based on the definition of upper shelf energy given in ASTM E185-82.(d) Surveillance Weld was fabricated from weld wire type SMIT 40, Heat # 25295, Flux type SMIT 89,Lot # 1103.13 PRESSURE TEMPERATURE LIMITS REPORTTable 5-2Calculation of Chemistry Factors using Sequoyah Unit 1 Surveillance Capsule DataMaterial Capsule Capsule f~a) FF) ARTNBT(C) FF*ARTNDT FF2Lower Shell T 2.61E+18 0.63 67.52°F 42.54°F 0.40Forging 04 U 7.96E+18 0.94 109.7°F 103.12°F 0.88(Tangential)X 1.32E+19 1.08. 145.12°F 156.73°F 1.16(Heat # 980919/ /_________281587) Y 2.19E+19 1.21 129.87°F 157.14°F 1.47Lower Shell T 2.6 1E+I18 0.63 50.59°F 31.87°F 0.40Forging 04 U 7.96E+18 0.94 67.59°F 63.53°F 0.88(Axial) X 1.32E+19 1.08 103.34°F 111.61°F 1.16(Heat #980919 / *Y 2.19E+19 1.21 133.35°F 161.35°F 1.47281587)SUM: 827.89°F 7.82CF04 =

  • RTNDTr) + 2( FF2) = (827.89) +(7.82) = 105.9°FSurveillance Weld T 2.61E+18 0.63 115.0°F 72.5°F 0.40Material(d) U 7,96E+18 0.94 130.4°F 122.6°F 0.88(Heat # 25295)(e) X 1.32E+19 1.08 143.1°F 154.5°F 1.16Y 2.19E+19 1.21 147.4OF 178.40F 1.47SUM: 528.0°F 3.91CF Surv. we~d =XY(FF
  • RTrrDT) +X( FF2) =(528.0°F) + (3.91) = 135.0°FNotes:(a) f= Calculated fluence from Capsule Y dosimetry analysis resultsETI, (n/cm2, E > 1.0 MeV).(b) FF = fluence factor =fI0.28"0.1logf.(c) ARTNDTValues are the measured 30 ft-lb shift values taken from App. B of Ref. 7, rounded to onedecimal point.(d) The surveillance weld metal ARTNDoT values have been adjusted by a ratio factor of 0.90.(e) Surveillance Weld was fabricated from weld wire type SMIT 40, Heat # 25295, Flux type SMIT 89,Lotft 110314 PRESSURE TEMPERATURE LIMITS REPORTTable 5-3Reactor Vessel Beltline Material Unirradiated Toughness Properties for Sequoyah Unit 1Material Description Cu (%) Ni (%) Initial RTNDT(a)Intermediate Shell Forging 05(Heat'#980807/281489) 01 .640Lower Shell Forging 04(Heat #980919/281587) 01 .670Surveillance Weld (Heat # 25295)(b'd, e) == 0.387 0.11 ---Rotterdam Test(c. e) 0.30 ......-Rotterdam Test(c' e) 0.25 ......Rotterdam Test(c' e) 0.46 ......-Best Estimate of the Intermediate to Lower ShellForging Circumferential Weld Seam W05 0.35 0.11 -40OF(Heat # 25295)(d. e)(a) The Initial RTNDT values are measured values(b) These copper and nickel values are best estimate values for only the surveillance weld metal and is the averageof three data points [0.424 (WCAP-10340, Rev.1), 0.406 (WCAP-10340, Rev.1), 0.33 (WCAP-8233) copperand 0.084 (WCAP-10340, Rev.1), 0.085 (WCAP-10340, Rev.1), 0.17 (WCAP-8233) nickel.]. These values aretreated as one data point in the calculation of the best estimate average for the inter, to lower shell circ. weldshown above. Originally the 0.424 / 0.406 and 0.084 / 0.085 values were reported as single points, 0.41 -0.42and 0.08 (Per WCAP-10340, Rev. l[7d]), but it is actually made up of two data points. Sample TW58 fromCapsule T was broken into two samples, TW58a and TW58b, thus providing the two data points.(c) From NRC Reactor Vessel Integrity Database (RVID) and ultimately fr'om Rotterdam Weld Certifications.(d) Circumferential Weld Seam W05 was fabricated with weld wire type SMvIT 40, Heat # 25295, Flux type SMVIT89, Lot # 2275. The surveillance weld was fabricated with weld Wire type SMIT 40, Heat # 25295, Flux typeSMIT 89, Lot # 1103 and is representative of the intermediate to lower shell circumferential weld.(e) The surveillance weld and the three Rotterdam tests are averaged together for the Best Estimate of theIntermediate to Lower Shell Forging Circumferential Weld Seam.15 PRESSURE TEMPERATURE LIMITS REPORTTable 5-4Peak Neutron Fluence Projections at Key Azimuthal Locations on the Reactor Vessel Clad/Base MetalInterface for Sequoyah Unit 1 (x 10 9 n/cm2, E > 1.0 MeV)Azimuthal LocationEFPY 00 150 300 45010.03 0.205 0.321 0.409 0.63720 0.387 0.596 0.761 1.1832 0.605 0.928 1.19 1.8448 0.896 1.37 1.75 2.7216 PRESSURE TEMPERATURE LIMITS REPORTTable 5-5Sequoyah Unit 1 Calculation of the ART Values for the 1/4T Location @ 32 EFPY(a)Material RG 1:99 CF FF ARTnDT(C) Margin(d) ART(e)R2 Method (0F) (°F) (0F) (0F) (0F)Intermediate Shell Forging 05 Position 1.1 115.6 1.029 40 119.0 34 193Position 1.1 95 1.029 73 97.8 34 205Lower Shell Forging 04Position 2.1 105.9 1.029 73 109.0 34(0 216Intermediate to Lower Shell Position 1.1 161.3 1.029 -40 166.0 56 182Circumferential Weld Seam Position 2.1 135.0 1.029 -40 138.9 56(0 155Notes:(a) Neutron fluence value used for all materials is the highest value from Table 5-4 for 32 EFPY.(b) Initial RTNIJT values are measured values.(c) AXRTNDT = CF
  • FF(d) M = 2 *(a'2 +- OA2)1/2(e) ART = Initial RTNDT + ARTNDT + Margin (0F)(f) Data deemed not-credible (See Reference 7a), thus the full GA will be used to determine margin.Table 5-6Sequoyah Unit 1 Calculation of the ART Values for the 3/4T Location @ 32 EFPY(a)Material RG 1.99 CF FF IRTrmT(b) ART~rTC() Margin(d) ART(e)*R2 Method (0F) (0F) (0F) (0F) (0F)Intermediate Shell Forging 05 Position 1.1 115.6 0.747 40 86.4 34 160Loe hl ogn 4 Position 1.1 *95 0.747' 73 71.0 34 178*Position 2.1 105.9 0.747 73 79.1
  • 34(0 186Intermediate to Lower Shell Position 1.1 161.3 0.747 -40 120.5 56 137Circumferential Weld Seam Position 2.1
  • 135.0 0.747 -40 100.8 56(0 117Notes:(a) Neutron fluence value used for all materials is the highest value from Table 5-4 for 32 EFPY.(b) Initial RTNDT values are measured values.(c) ARTrDT =CF *FF(d) M = 2 *(a.2 + Gra2)1t2(e) ART = Initial RTNDT .+- ARTrNDT + Margin (0F)(f) Data deemed not-credible (See Reference 7a), thus the full GrA will be used to determine margin.17 PRESSURE TEMPERATURE LIMITS REPORTTable 5-7Summnary of the Sequoyah Unit 1 Reactor Vessel Beitline Material ART ValuesMaterial RG 1.99 R2 1/4 ART 3/4 ARTMethod (0F) (0F)Intermediate Shell Forging 05 Position 1.1 193 160Position 1.1 205 178Lower Shell Forging 04Position 2.1 216 186Intermediate to Lower Shell Position 1.1 182 137Circumferential Weld Seam Position 2.1 155 117Table 5-8RTP-s Calculations for Sequoyah Unit 1 Beltline .Region Materials at 32 EFPY(a)Material Fluence FF CF ARTpTs~b) Margin RTpxs(d)(X 1019 n/cm2, "(F) (0F) (0F) (OF) (0F)E>1.0 MeV)Intermediate Shell Forging 05 1.84 1.167 115.6 .134.9 34 40 209Lower Shell Forging 04 1.84 1.167 95.0 110.9 34 73 218Lower Shell Forging 04 1:84 1.167 105.9 123.6 34(e) 73 231(Using S/C Data)Circumferential Weld Metal 1.84 1.167 161.3 188:2 56 -40 204Circumferential Weld Metal 1.84 1.167 135.0 157.5 56(e) -40 174(Using S/C Data)Notes:(a) Neutron fluence value used for all materials is the highest value from Table 5-4 for 32 EFPY.(b) ARTPTs = CF *FF(c) Initial RTNOT values are measured values(d) RTPTs= + ARTpTs + Margin (0F)(e) Data deemed not-credible (See Reference 7a), thus the full will be used to determine margin.18 PRESSURE TEMPERATURE LIMITS REPORT6.0 References1. WCAP-8233, Tennessee Valley Authority Sequoyah Unit No. 1 Reactor Vessel RadiationSurveillance Progr~am, S. E. Yanichko, et. al., December 1973.2. Code of Federal Regulations, 10OCFR50, Appendix H, Reactor Vessel Material SurveillancePro gramn Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C..3. ASTM E23 Standard Test Method Notched Bar himpact Testing of Metallic Materials, in ASTMStandards, American Society for Testing and Materials, Philadelphia, PA.4.Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G, Fracture Toughness"Criteria for Protection Against Failure5. ASTM E185-82, Annual Book of ASTM Standards, Section 12, Volume 12.02, Standard Practicefor Conducting Surveillance Tests for Light- Water Cooled Nuclear Power Reactor Vessels.6. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S.Nuclear Regulatory Commission, May 1988.7a. WCAP-15224, Analysis of Capsule the Tennessee Valley Author"ity Sequoyah Unit 1Reactor Vessel Radiation Surveillance Program, T.J. Laubham, et. al., June 1999.7b. WCAP-1 3333, Analysis of Capsule Xfoino the Tennessee Valley Authority Sequoyah Unit 1Reactor Vessel Radiation Surveillance Progr~am, M.A. Ramirez, S. L. Anderson, L. Albertin, June1992.7c. .SwRI Project 06-8851, Reactor Vessel Material Surveillance Progr'am for Sequoyah Unit No. 1:Analysis of Capsule U, P. K. Nair, et al., October 1986.7d. WCAP- 10340, Revision 1, Analysis of Capsule T foin the Tennessee Valley Authority SequoyahUnit.] Reactor Vessel Radiation Surveillance Program, S.E. Yanichko, et. al., February 1984.8. CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.1, developed by ATIConsulting, March 1999.9. WCAP-14040-NP-A, Revision 2, Methodology Used to Develop Cold Overpressure MitigatingSystem Setpoints and RCS Heatup and Cooldown Limit Curves, J.D. Andrachek, et. al., January1996.10. WCAP-15293, Revision 2, Sequoyah Unit 1 Heatup and Cooldown Limit Curves for NormalOperation and PTLR Support Documentation, J.H. Ledger, July 2003.11. Westinghouse Letter to TVA, TVA-93 -105, Cold Overpressure Mitigation System Code Case andDelta-P Calculation, dated May 19, 1993.12. Calculation SQN-IC-01 4, Demonstrated Accuracy Calculation for Cold Overpressure ProtectionSystem.19 PRESSURE TEMPERATURE LIMITS REPORT13. ASME Code Case N-640, Alternative Reference Fracture Toughness for Development of P-TLimit Curves for Section XL, Division 1, dated February 26, 1999.14. WCAP-15984-P, Revision 01, Reactor Vessel Closure Head/Vessel Flange RequirementsEvaluation for Sequoyah Units 1 and 2, W. Bamford, et.al., April 2003.20 ENCLOSURE 3SEQUOYAH UNIT 2PRESSURE TEMPERATURE LIMITS REPORT, REVISION 6 Bi88 151.016"PRESSURE TEMPERATURE LIMITs REPORT801IContractrfo ny port of his re-.dligdealsad dinmsnsions. .*: 'Octobe~r 16,.2015. _* SOEP (Nt) BYW. J.Pierce__*Tennessee Valley AuthoritySequoyah unit 2Pressure Temperature Limits ReportRevision 6, september 2015PROJECT Secjuovah 'DisCIPLINE NCONTRACT " 4411 UNIT. 2DESC. RCS Pressure-Temperature L'mait-ReportDWG/DOC NO. PTLR-2SHEET ..- OF -REV. 06*DATE. 10/16/15 ECN/DCN FILE N2N-081"EDMS, WT CA-K PRESSURE TEMPERATURE LIMITS REPORTTable of ContentsLito T b e .List. .................of.............Tables........... .......................iList of Figures ............................................................................................. ........ v1.0 RCS Pressure Temperature Limits RePort (PTLR) ........ .... .................................... 12.0 Operating Limits ............................................................................. .............12.1 RCS Pressure/Temperature (PIT) Limits (TS 3.4.3)................................................. 13.0 Low Temperature Overpressure Protection System (TS 3.4.12) ...................... ..13.1 Pressurizer PORV Lift Setting Limits ........... .....................3.2. Arming Temperature ............'......24.0 Reactor Vessel Material Surveillance Program.....,................................................. 25.0 Supplemental Data Tables ............................................................................3..36.0 References ......................................... .. ... ..... ........ 17ii PRESSURE TEMPERATURE LIMITS REPORTList of TablesTable 2-1 Sequoyah Unit 2 Heatup Limits at 32 BEFPY(with Uncertainties for Instrumentation Errors of l0°F and 60 psig)............................... 6Table 2-2 Sequoyah Unit 2 Cooldown Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 10°F and 60 psig)...... ......................... 7Table 3.-1 Selected Setpoints, Sequoyah Unit 2........................................................... 8Table 4-1 Sequoyah Unit 2 Reactor Vessel Surveillance Capsule Withdrawal Schedule............. 10Table 5-1 Comparison of the Sequoyah Unit 2 Surveillance Material 30 ft-lb TransitionTemperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99,revision 2, Predictions................ .......................................................... 11Table 5-2 Calculation of Chemistry Factors using Sequoyah Unit 2 Surveillance Capsule Data....12Table 5-3 Reactor Vessel Beltline Material Unirradiated Toughness Properties forSequoyah Unit2 ................................................................................ 13Table 5-4 Peak Neutron Pluence Projections at Key Azimuthal Locations on the Reactor VesselClad/Base Metal Interface for Sequoyah Unit 2 (x 1019 n/cm2, EB> 1.0 MeV) ....i........ 14Table 5-5 Sequoyah Unit 2 Calculation of the ART Values for the 1/4T Location @ 32 EFPY.....15Table 5-6 Sequoyah Unit 2 Calculation of the ART Values for the 3/4T Location @ 32 EPPY.....15Table 5-7 Summary of the Limiting ART Values Used in the Generation of the Sequoyah Unit 2Heatup/Cooldown Curves.......................................................... :............ 16Table 5-8 RTPTs Calculations for Sequoyah Unit 2 Beltline Region Materials at 32 EFPY.......... 16iii PRESSURE TEMPERATURE LIMITS REPORTList of FiguresFigure 2-1 Sequoyah Unit 2 Reactor Coolant System Heatup Limitations(Heatup Rate of I 00°F/hr) Applicable for the First 32 EFPY(wi/Margins for'Instrumentation Errors of 1 00F and 60 psig)................................ 4Figure 2-2 Sequoyah Unit 2 Reactor Coolant System Cooldown Limitations(Cooldown Rates up to 1 00°F/hr) Applicable for the First 32 EFPY(w/ Margins for Instrumentation Errors of 10°F and 60 psig) .................... .............5Figure 3-1 Sequoyah Unit 2 Selected LTOPS Setpoints .................................................. 9iv PRESSURE TEMPERATURE LIMITS REPORT1.0 RCS Pressure Temperature Limits Report (PTLR)This PTLR for Sequoyah Unit 2 has been prepared in accordance with the requirements of TechnicalSpecification (TS) 5.6.4. Revisions to the PTLR shall be provided to the NRC after issuance.This report affects TS.3.4.3, RCS Pressure/Temperature Limits (P/T) Limits and TS 3.4.12,.LowTemperature Overpressure Protection (LTOP) System.2.0 RCS Pressure and Temperature LimitsThe limits for TS 3.4.3 are presented in the subsections which follow and were developed using the NRCapproved methodologies specified in TS 5.6.4 with exception of ASME Code Case N-64011n1 (Use of K1o),WCAP- 159 84-Ptl21 (Elimination of the Flange Requirement), 1996 Version of Appendix G[41 and therevised fluencesETI. The operability requirements associated with LTOPS are specified in TS 3.4.12 andwere determined to adequately protect the RCS against brittle fracture in the event of an LTOP Transientin accordance with the methodology specified in TS 5.6.4.2.1 RCS Pressure/Temperature (P/T) Limits (TS 3.4.3)2.1.1 The minimum boltup temperature is 50°F2.1.2 The RCS temperature rate-of-change limits are:a. A maximum heatup rate of 1 00°F in any one hour period.b. A maximum cooldown rate of 100°F in any one hour period.c. A maximum temperature change of less than or equal to. 1 0°F in any one hour period duringinservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.2.1.3 The RCS P/T limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticalityare specified by Figures 2-1 and 2-2.3.0 Low Temperature Overpressure Protection System (TS 3.4.12)The lift setpoints for the pressurizer Power Operated Relief Valves (PORVs) are presented in thesubsections which follow. These lift setpoints have been developed using the NRC-approvedmethodologies specified in TS 5.6.4.I PRESSURE TEMPERATURE LIMITS REPORT*3.1 Pressurizer PORV Lift Setting LimitsThe pressurizer PORV lift setpoints are specified by Figure 3-1 and Table 3-1 (Ref. 10). The limits for theLTOPS setpoints are contained in the 32 EFPY steady-state curves (Table 2-2), which are beitlineconditions and are not compensated for pressure differences between the pressurizer transmitter and thereactor midplane/beltline or for instrument inaccuracies. The pressure difference between the pressurizertransmitter and the reactor vessel midplane/beltline with four reactor coolant pumps in operation is 68.3psi (Ref. 13).Note: These set-points include allowance for the pressure difference between the pressurizertransmitter and the reactor vessel midplane/beltline and the 50°F thermal transport effect forheat injection transients.. A demonstrated accuracy calculation (Reference 14) has beenperformed to confirm that the setpoints will maintain the system pressure within theestablished limits when the pressure difference between the pressure transmhitter and reactormidplane and maximum temperature/pressure instrument uncertainties are applied to thesetpoints.3.2 Arming TemperatureThe LTOPS arning temperature is based upon the methodology defined in the Sequoyah Nuclear PlantUnit 2 Technical Specifications Administrative Controls Section 5.6.4. The arming temperature shall be <350°F.4.0 Reactor Vessel Material Surveillance ProgramThe reactor vessel material irradiation surveillance specimens shall be removed and examined todetermine changes in material properties. The removal schedule is provided in Table 4-1. The results ofthese examinations shall be used to update Figures 2-1, 2-2 and 3-1.The pressure vessel steel surveillance program (WCAP-85 13[1]) is in compliance with Appendix H to 10CFR 50, "Reactor Vessel Material Surveillance Program Requirementst21."' The material test requirementsand the acceptance standard utilize the reference nil-ductility temperature RTNDT, which is determined inaccordance with ASTM E23 [3. The empirical relationship between RTNOT and the fracture toughness ofthe reactor vessel steel is developed in accordance with Code Case N-640 of Section XI of the ASMEBoiler and Pressure Vessel Code, Appendix G, "Fracture Toughness Criteria for Protection AgainstFailurer4k.' The surveillance capsule removal schedule meets the requirements of ASTM E185-82N5.Theremoval schedule is provided in Table 4-1.2 PRESSURE TEMPERATURE LIMITS REPORT5.0 Supplemental Data TablesTable 5-1 contains a comparison of measured surveillance material 30 ft-lb transition temperature shiftsand upper shelf energy decreases with Regulatory Guide 1.99, Revision 2[6], predictions.Table 5-2 shaows calculations of the surveillance material chemistry factors using surveillance capsuledata. Note that in the calculation of the surveillance weld chemistry factor, the ratio procedure fromRegulatory Guide 1.99, Revision 2 was followed. The ratio in question is equal to 0.93.Table 5-3 provides the required Sequoyah Unit 2 reactor vessel toughness data.Table 5-4 provides a summary of the fluence values used in the generation of the heatup and cooldownlimit curves and the PTS evaluation.Table 5-5 and 5-6 show the calculation of the 1/4T and 3/4T adjusted reference temperature at 32 EFPYfor each beltline material in the Sequoyah Unit 2 reactor vessel. The limiting beltline material was theIntermediate Shell Forging 05..Table 5-7 provides a summary of the adjusted reference temperature (ART) values of the Sequoyah Unit 2reactor vessel beltline materials at the l/4T and 3/4T locations for 32 EFPY.Table 5-8 provides RTP-Ts values for Sequoyah Unit 2 at 32 EFPY.3 PRESSURE .TEMPERATURE LIMITS REPORTMATERIAL PROPERTY BASISLIMITING MATERIAL! INTERMEDIATE SHELL FORGING 05LIMITING ART VALUES AT 32 EFPY: 1/4T, 142°F3/4T, l15°F2Rflfl-v Io)perlim Version:5.1 Run:5694 /ILeak, Test, Limit 22502000--[ Unacceptable1Oeain ....*Acceptable ...Operation1750 -..1500125U) 00Co0 25~ritical Limit*100 Deg. F/Hr750"500250MinumumBoltup Temp-- = 500F "Criticality Limit based on-~inservic e hydrostatic testtemperature (214°F) for theservice period up to 32 EFPYUf.......I ..............! ....! ........ .........0 50 100 150 200 250 300 350 400 450 500 550Moderator Temperature (Deg. F)Figure 2-1 Sequoyah Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of1000F/hr) Applicable for the First 32 EFFY (w/ Margins for InstrumentationError of 10°F and 60 psig) (Plotted Data provided on Table 2-1)

PRESSURE TEMPERATURE LIMITS REPORTMATERIAL PROPERTY BASISLIMITING MATERIAL: INTERMEDIATE SHELL FORGING 05.LIMITING ART VALUES AT 32 EFPY: 1/4T, 142°F3/4T, 150F.2500 [Oeri Version:5.1 Run:5694]2000 tUnacceptable _____ --__ _________ AcceptableOperation OPeration -1750 ..... 'S 1250 * -.-o----__ -__ __ ____ '--"o Cooldown~~Rates i"= 1000 F/Hr .......................u steady-stateo-6o1750 60 oo250 -.... ...........M inim um I- ......:.. ........ .. ... ...4.. ... .........SBoltup Templi ..! ' i-500F !0 50 100 150 200 250 300 350 400 450 500 550Moderator Temperature (Deg. F)Figure 2-2 Sequoyahi~nit 2 Reactor Coolant System Cooldown Limitations (CooldownRates up to 100°F/hr) Applicable for the First 32 EFPY (w/ Margins forInstrumentation Error of 100F and 60 psig) (Plotted Data provided on Table 2-2)2 PRESSURE TEMPERATURE LIMITS REPORTTable 2-1Sequoyah Unit 2 Heatup Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 10°F and 60 psig)100 Heatup 1100 Critical Limit ILeak Test LimitT P T P JT P50 0 214 0 198 200050 591 214 607 214 248555 595 214 61460 601 214 62265 607 214 65770 614 214 65075 622 214 64780 630 214 64685 640 214 64890 646 214 *65395 646 214 661100 6.46 214 671105 646 214 680110 646 214 685115 646 214 701120 646 214 720125 648 214 743130 653 214 769135 661 214 798140 671 215 832145 685 220 869150 701 225 911155 720 230 959160. 743 235 1011165 769 240 1069170 798 245 1134175 832 250 1206180 869 255 1286185 911 260 1374190 959 265 1471195 1011 270 1579200 1069 275 1698205 1134 280 1829210 1206 285 1974215 1286 290 2134220 1374 295 2311225 1471230 1579235 1698240 1829245 1974250 2134255 23113 PRESSURE TEMPERATURE LIMITS REPORTTable 2-2Sequoyah Unit 2 Cooldown Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 1 0°F and 60 psig)Steady State 1 20F 40F 60F I00OFT P T P T P T P T P5050556065707580859095100105110115120125130135140145*1501551601651761751801851901952002052102152202252300591595601607614622630640650661674688703720739.7607838098378689029409821028108011361199126813441429152216251739186520042158232850505560657075808590951001051101151201251301351401451501551601650552554558564572580589599610623636652668687707730755783814848885927.973102450505560657075808590951001051101151201251301351401451501551600503508514521529538548*55957158459961663465467670172975979383187291896850505560657075808590951001051101151201251301351401451501550.461466470478486496506518531546562580600622647674704738775816862913505055606570758085909510010511011512012513013514014515003663723803893994104234374534704905125365635936266637047498008564 PRESSURE TEMPERATURE LIMITS REPORTTable 3-1Selected Setpoints, Sequoyah Unit 2PORV#2 PORV#1Trcs (Deg.F) Setpoint (psig) Setpoint (psig)50 510 485100 580 555135 640 610174 745 682200 745 685*250 745 *685-278 745 685400 745 685450 2350 23505 PRESSURE TEMPERATURE LIMITS REPORTSequoyah Unit 2 LTOPS Selected Setpoints-/! [4-2500 --....J Lo- I I !0 01015 0 5 303040 5 0Reacto Coln Iyte enpraue(FA F--*-- FORV#2-FinaI --U-- FORV#1Figure 3-1 Sequoyah Unit 2 Selected LTOPS Setpoints (Plotted Data pro vided on Table 3-1)6 PRESSURE TEMPERATURE LIMITS REPORTTable 4-1Sequoyah Unit 2 Reactor Vessel Surveillance Capsule Withdrawal Schedule(a) Updated in Capsule Y dosimetry analysis (WCAP-15320t71).(b) Effective Full Power Years (EFPY) from plant startup.(c) Plant specific evaluation.(d) This fluence is not less than once or greater than twi ce the peak end of license (32 EFPY) fluence(e) Capsules 5, V, W and Z will reach a fluence of 2.71 x 1019 (EB> 1.0 MeV), the 48 EFPY peakvessel fluence at approximately 44 EFPY.,Administrative Note -The surveillance capsule withdrawal schedule in Table 4-1 is based on thesurveillance program for the original 40 year service life. Relocation of select standby, capsules toincrease the fluence lead factor in anticipation of the updated surveillance program for the 60 year licenserenewal service life is described in a TVA letter to NRC dated January 10, 2013 (ML13032A25 1). .Regulatory approval for the anticipatory standby capsule relocation has been granted by a NRC letter toTVA dated September 27, 2013 (ML13240A320). A complete update of the reactor vessel surveillanceprogram for the 60 year license renewal service life will be documented by a subsequent revision to thePTLR prior to entry into the license renewal extended operating period.7 PRESSURE TEMPERATURE LIMITS REPORTTable 5-1Comparison of the Sequoyah Unit 2 Surveillance. Material 30 ft-lb Transition Temperature Shifts andUpper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions30 ft-lb Transition Upper Shelf EnergyTemperature Shift DecreaseMaterial Capsule Fluence ' Predicted Measured Predicted Measured(x 1019 n/cm2) (OF)(a) (OF)(b) (%,/)(al) (%)(c)Intermediate Shell T 0.261 60.33. 63.65 17 12Forging 05 U 0.692 85.22 79.31 21 16(Tangential)(et285/X 1.22 100.23 85.7 23 8981057) Y 2.14 114.67 134.12 26 22Intermediate Shell T 0.261 60.33 48.73 17 7Forging 05 U 0.692 85.22 66.06 21 9(Axial)(et285/X 1.22 100.23 110.04 23 2981057) Y 2.14 .114.67 89.21 26 22Weld Metal T 0.261 43.12 74.56 20 2(Heat # 4278)(d) U 0.692 60.91 130.38 25 6X 1.22 71.63 44.22 29 35Y 2.14 81.96 86.91 33 "3HAZ Metal T 0.261 --24.58 -,- 2U 0.692 --64.03 --14X 1.22 --28.29 --19Y 2.14 --50.32 --39Notes:(a) Based on Regulatory Guide 1.99, Revision 2, methodology using the mean weight percent values ofcopper 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) Surveillance Weld was fabricated from weld wire type SMIT 89, Heat # 4278, Flux type SMIT 89,Lot # 1211.8 PRESSURE TEMPERATURE LIMITS REPORTTable 5-2Calculation of Chemistry Factors using Sequoyah Unit 2 Surveillance Capsule DataMaterial Capsule Capsule f(a) FF(b) ARTNDT(C) FF*ARTNDT FF2Intermediate Shell T 2.61E+18 0.635 63.7 40.45 0.403Forging 05 U 6.92E+18 0.897 79.3 71.13 0.805(Tangential) X 1.22E+19 1.055 85.7 90.41 1.113(Heat #288757 / Y 2.14E+19 1.207 134.1 161.86 1.457981057)Intermediate Shell T 2.61E+18 0.635 48.7 30.92 0.403Forging 05 U 6.92E+18 0.897 66.1 -59.29 0.805(Axial) X 1.22E+19 1.055 110.0 116.05 1.113(Heat #288757 / Y 2.14E+19 1.207 89.2 107.66 1.457981057) ______ ______SUM: 677.77°F 7.556CFo5 = X(FF

  • RTNDT) + X.( FF2) = (677.77) +(7.556) = 89.70FSurveillance Weld T 2.6 1E+18 0.635 69.4 (74.6) 44.07 0.403Material(d) *U 6.92E+18 0.897 121.3 (130.4) 108.81 0.805(Heat # 4278)(e) X 1.22E+19 1.055 41.1 (44.2) 43.36 1.113Y 2.14E+19 1.207 80.8 (86.9) 97.53 1.457SUM: 293 .77°F 3.778CF Surv. weld =
  • RTNDT) + X( FF2) = (293.77°F) +(3.778) = 77.8"FNotes:(a) f = Calculated fluence from Capsule Y dosimetry analysis results [7], (n/cma2, E > 1.0 MeV).(b) FF = fluence factor = f02-.~o 3(c) AIRTNDT values are the measured 30 ft-lb shift values taken from App. B of Ref. 7, rounded to onedecimal point.(d) The surveillance weld metal ARTNDT Values have been adjusted by a ratio factor of 0.93.(e) Surveillance Weld was fabricated from weld wire type SMIT 89, Heat # 4278, Flux type SMIT 89, Lot #1211.

PRESSURE TEMPERATURE LIMITS REPORTTable 5-3Reactor Vessel Beltline Material Unirradiated TouglmessProperties for Sequoyah Unit 2Material Description Cu (%) Ni (%) Initial RTNDT(a)Intermediate Shell Forging 05(Heat #288757 /981057)013.710FLower Shell Forging 040.14 0.76 -22 °F(Heat # 990469 / 293323)Intermediate to Lower Shell ForgingCircumferential Weld Seam W05(b) 0.12 0.11 -40F(Heat # 4278)Surveillance Weld(b) 0.13 0.11Notes:(a) The Initial RTNDT values are measured values(b) Circumferential Weld Seam was fabricated with weld wire type SMIT 89, Heat # 4278, Flux type SMIT89, Lot # 1211 and is representative of the intermediate to lower shell circumferential weld.10 PRESSURE TEMPERATURE LIMITS REPORTTable 5-4Peak Neutron Fluence Projections at Key Azimuthal Locations on the Reactor Vessel Clad/Base MetalInterface for Sequoyah Unit 2 (x 1019 n/cm2, E > 1.0 MeV)Azimuthal LocationEFPY 0° 150 3 00 45010.54 0.211 0.336 0.426 0.63720 0.38 0.60 0.773 1.1632 0.593 0.934 1.21 1.8248 0.878 1.38 1.80 2.7111 PRESSURE TEMPERATURE LIMITS REPORTTable 5-5Sequoyah Unit 2 Calculation of the ART Values for the 1/4T Location @ 32 EFPY(a)Material RG 1.99 cF FF IRTNDT(b) ARTNDTo(C) Margin(d) ART(e)R2 Method (0F) _____ (0F) (0F) (0F) (0F)Position 1.1 95 1.027 10 97.6 34 142Intermediate Shell Forging 05Position 2.1 89.7 1.027 10 92.1 34 136Lower Shell Forging 04 Position 1.1 104 1.027 -22 106.8 34(0 119Intermediate to Lower Shell " Position 1.1 .63 '1.027 -4 64.7 56 117Circumferential Weld Seam Position 2.1 77.8 " 1.027 .-4 79.9 56(0 132Notes:(a) Neutron fluence value used for all materials is the highest value from Table 5-4 for 32 EFPY.(b) Initial RTNDoT values are measured values.(c) ARTrmT CF *FF(d) M =2 *(a'i2 +I 0'A2)1/2(e) ART =Initial RTNDT +- ARTNDTr + Margin (0F)(f) Data deemed not-credible (See Reference 7a), thus the full ca will be used to determine margin.Table 5-6Sequoyah Unit 2 Calculation of the ART Values for the 3/4T Location @ 32 EFPY(a)Material RG 1.99 CF FF IRTNDT(b) Margin(d) ART(e)R2 Method (0F) (0F) (0F) (0F) (0F)Position 1.1 95 0.745 10 70.8 34

  • 115Intermediate Shell Forging 05Position 2.1 89.7 0.745 10 66.8 34 .111Lower Shell Forging 04 Position 1.1 104 0.745 -22 77.5 34(0 90intermediate to Lower Shell Position 1.1 63 0.745 .-4 46.9 56 99Circumferential Weld Seam Position 2.1 77.8 0.745 -4 58.0 56(0 110Notes: .(a) Neutron fluence value used for all materials is the highest value from Table 5-4 for 32 EFPY.(b) Initial RTrNDT values are measured values.(c) =CF *FF(d) M = 2 *(0i2 + raA)2(e) ART =Initial RTrNDv + ARTNDT + Margin (0F)(f) Data deemed not-credible (See Reference 7a), thus the full GA will be used to determine margin.12 PRESSURE TEMPERATURE LIMITS REPORTTable 5-7Summary of the Sequoyah Unit 2 Reactor Vessel Beitline Material ART ValuesMaterial RG 1.99 R2 1/4 ART 3/4 ARTMethod (0F) (0F)Position 1.1 142 115Intermediate Shell Forging 05Position 2.1 136 111Lower Shell Forging 04 Position 1.1 119 90Intermediate to Lower Shell Position 1.1 117 99Circumferential Weld Seam Position 2.1I 132 110Table 5-8RTPTs Calculations for Sequoyah Unit 2 BeltlineRegion Materials at 32 EFPY(a)Material Fluence FF CF ARTpTs~b) Margin RTNDT(U-)(C) RTpTs(d)(x 10 19 n/cm2, (0F) (0F) (0F) (0F) (0F)E>l.0 MeV)Intermediate Shell Forging 05 1.82 1.164 95 110.6 34 10 155Intermediate Shell Forging 05 1.82 1.164 89.7 104.4 34 10 148(Using S/C Data) _______Lower Shell Forging 04 1.82 .1.164 104 121.1 34(e) -22 133Circumferential Weld Metal 1.82 1.164 63 73.3 56 -4 125Circumferential Weld Metal 1.82 1.164 77.8 90.6 56(e) -4 143(Using S/C Data)____Notes:(a) Neutron fluence value used for all materials is the highest value from Table 5-4 for 32 EFPY.(b) ARTpTs= CF
  • FF(c) Initial RTNDT values are measured values(d) RTpTs = RTNDT(U) + ARTpTs + Margin (0F)(e) Data deemed not-credible (See Reference 7a), thus the full ra.. will be used to determine margin.13 PRESSURE TEMPERATURE LIMITS REPORT6.0 References1. WCAP-8513, Tennessee Valley Authority Sequoyah Unit No. 2 Reactor Vessel RadiationSurveillance Program, J. A. Davidson, et. al., November 1975.2. Code of Federal Regulations, 10OCFR50, Appendix H, Reactor Vessel Material SurveillanceProgr~am Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C.3. ASTM E23 Standard Test Method Notched Bar" Impact Testing of Metallic Materials, in ASTMStandards, American Society for Testing and Materials, Philadelphia, PA.4.Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G, Fracture ToughnessCriteria for Protection Against Failure5. ASTM E185-82, Annual Book of ASTM Standards, Section 12, Volume 12.02, Standard Practicefor Conducting Surveillance Tests for Light- Water Cooled Nuclear Power Reactor Vessels.6. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S.Nuclear Regulatory Commission, May 1988.7a. WCAP- 15320, Analysis of Capsule Y fi'omn the Tennessee Valley Authority Sequoyah Unit 2Reactor Vessel Radiation Surveillance Program, T.J. Laubham, et. al., November 1999.7b. WCAP-*10509, Analysis of Capsule T fiom the. Tennessee Valley Authority Sequoyah Unit 2Reactor Vessel Radiation Surveillance Program, R. S. Boggs, et al, April 1984.7c. Southwest Research Institute Nondestructive Evaluation Science and Technology Division,Reactor Vessel Material Surveillance Program and Technology Division, Reactor Vessel Material"Surveillance Pro gram for Sequoyah Unit 2." Analysis of Capsule U, Final Report SwRI Project*17-8851 TVA Contra~ct 85PJH-964430, January !1990.7d. WCAP- 13545, Analysis of Capsule X from the Tennessee Valley Authority Sequoyah Unit 2Reactor Vessel Radiation Surveillance Program, M. A. Ramirez, S. L. Anderson, A. Madeyski,November 1992.8. CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.1, developed by ATIConsulting, March 1999.9. WCAP- 14040-NP-A, Revision 2, Methodology Used to Develop Cold Overpressure MitigatingSystem Setpoints and RCS Heatup and Cooldown Limit Curves, J.D. Andrachek, et. al., January1996.10. WCAP-15321, Revision 2, Sequoyah Unit 2 Heatup and Cooldown Limit Curves for NormalOperation and PlTLR Support Documentation, J.H. Ledger, et.al., July 2003.11. ASME Code Case N-640, Alternative Reference Fracture Toughness for Development of P-T*Limit Curves for Section XL, Division 1, dated February 26, 1999.14 PRESSURE TEMPERATURE LIMITS REPORT12. WCAP-15984-P, Revision 01, Reactor Vessel Closure Head/Vessel Flange RequiremnentsEvaluation For Sequoyah Units 1 and 2, W. Bamford, et.al., April 2003.13. Westinghouse Letter to TVA, TVA-93-105, Cold Overpressure Mitigation System Code Case andDelta-P Calculation, dated May 19, 1993.14. Calculation SQN-IC-0 14, Demonstrated Accuracy Calculation for Cold Overpressure ProtectionSystem.15 Tennessee Valley Authority, Post Office Box 2000, Soddy Daisy, Tennessee 37384-2000November 13, 201510 CFR 50.4ATT-N: Document Control DeskU.S. Nuclear Regulatory CommissionWashington, D.C. 20555-0001Sequoyah Nuclear Plant, Units I and 2Renewed Facility Operating License Nos. DPR-77 and DPR-79NRC Docket Nos. 50-327 and 50-328

Subject:

Sequoyah Unit I Pressure Temperature Limits Report,Revision 5, and Sequoyah Unit 2 Pressure Temperature LimitsReport, Revision 6

References:

1. Letter from NRC to TVA, "Sequoyah Nuclear Plant, Units I and 2-Issuance of Amendments for the Conversion to the ImprovedTechnical Specifications with Beyond Scope Issues (TAC Nos.MF3128 and MF3129)," dated September 30, 20015(MLI15238B460)In accordance with Sequoyah Nuclear Plant (SQN) Units 1 and 2 TechnicalSpecifications (Tss) 5.6.4.c, enclosed is the Unit I Pressure Temperature LimitsReport (PTLR), Revision 5, and Unit 2 PTLR, Revision 6. In accordance with TSs5.6.4.c, the PTLRs are required to be provided to the Nuclear RegulatoryCommission (NRC) within 30 days after any revision. Sequoyah Units I and 2 wereissued license amendment Nos. 334 and 327, respectively for improved standardTSs (Reference 1). These license amendments resulted in the enclosed revisions toeach of the PTLRs. The revisions also include other editorial clarifications andadministrative changes identified during the revision process as described inEnclosure 1. The revised PTLRs became effective on October 16, 2016.There are no new regulatory commitments in this letter. If you have any questions,please contact Jonathan Johnson, SQN Site Licensing Manager at (423) 843-8129.

U.S. Nuclear Regulatory CommissionPage 2November 13, 2015Sequo Nuclear PlantEnclosures1.2.3.Units 1 and 2 Pressure Temperature Limits Report ChangesSequoyah Unit 1 Pressure Temperature Limits Report, Revision 5Sequoyah Unit 2 Pressure Temperature Limits Report, Revision 6ZTK: DVGEnclosurescc (Enclosures):NRC Regional Administrator -Region IINRC Senior Resident Inspector -SQN ENCLOSURE1ISEQUOYAH UNITS 1 AND 2PRESSURE TEMPERATURE LIMITS REPORT CHANGESThe following describes the editorial clarifications and administrative changes madeto each Units' Pressure Temperature Limits Report (PTLR).1. Conflict between the figure index referencing Cold Overpressure MitigationSystem (COMS) and the figure title referencing Low TemperatureOverpressure Protection System (LTOPS) was Corrected for consistency withthe system terminology in Section 3.4.12 of the Technical Specifications.2. Section 1, "RCS Pressure Temperature Limits Report (PTLR)," was revisedto clarify which Limiting Condition for Operations are affected by the PTLR.3. Section 3.1, "Pressurizer PORV Lift Setting Limits," was revised forconsistency with the analysis of record contained in Topical Report No.WCAP-1 5293, Revision 2 "Sequoyah Unit 1 Heatup and Cooldown LimitCurves for Normal Operation and PTLR Support Documentation," andTopical Report No. WCAP-15321 Revision 2 "Sequoyah Unit 2 Heatup andCooldown Limit Curves for Normal Operation and PTLR SupportDocumentation".4. Section 5.0, "Supplemental Data Tables," added the word "Forging" to thesentence in refer to the limiting beltline material identified in Tables 5-5 and5-6.5 A note was added to Table 4-1, "Sequoyah Unit 1 Reactor VesselSurveillance Capsule Withdrawal Schedule," to describe current and futureactivity with the surveillance capsule withdrawal schedule.6. Tables 5-5 and 5-6 were revised with a different annotation conventionconsistent with the other tables in the PTLR.

ENCLOSURE 2SEQUOYAH UNIT 1PRESSURE TEMPERATURE LIMITS REPORT, REVISION 5 B88 15 t 16800MAPRESSURE TEMPERATURE LIMITS REPORT....APPROVEDThis qp~provaIl does. not reliev, theContractor from any port o9 his re-,*ponsibility Jar, the correctness of.* design. debugailnd dimlnsio~ns.YLetter lo.Nl107l6.*oot,: ,October 16, 2015. -.SQEP BY: W. JX PierceTennessee Valley AuthoritySequoyah unit 1Pressure Temperature Limits ReportRevision 5, September 2015PROJECT Sequovah .DISCIPLINE N"CONTRACT .4411 1JNT' 1DESC. RCS Pressure-Temperature Limit Report*DWG/DOC NO. PTLR-1SHEET -OF -. "REV. 05*DATE 10/16/15 ECN/DCN -. FILE N2N-081EDMS, WT CA-K PRESSURE TEMPERATURE LIMITS REPORTTable of ContentsList of Tables .................................................................................................... ivList of Figures......................................................................................................v1.0 RCS Pressure Temperature Limits Report (PTLR) ......................................... ....... 12.0 Operating Limits.......................................................................................... 12.1 RCS Pressure/Temperature (PIT) Limits (TS 3.4.3) ............................................... 13.0 Low Temperature Overpressure Protection System (TS 3.4.12)...... ............................ 13.1 Pressurizer PORV Lift Setting Limits .............................................................. 23.2 Arming Temperature................................................................................ 24.0 Reactor Vessel Material Surveillance Program...................................................... 25.0 Supplemental Data Tables............................................................................... 36.0 References................................................................................................ 19ii PRESSURE TEMPERATURE LIMITS REPORTList of TablesTable 2-1 Sequoyah Unit 1 Heatup Limits at 32 EPPY(with Uncertainties for Instrumentation Errors of 10°F and 60 psig)......... ...................... 6Table 2-2 Sequoyah Unit 1 Cooldown Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 1 0°F and 60 psig)................................STable 3-1 Selected Setpoints, Sequoyah Unit 1.......................................................... 10Table 4-1 Sequoyah Unit I Reactor Vessel Surveillance Capsule Withdrawal Schedule ............ 12Table 5-1 Comparison of the Sequoyah Unit 1 Surveillance Material 30 ft-lb TransitionTemperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99,revision 2, Predictions .............................'............................................. 13Table 5-2 Calculation of Chemistry Factors using Sequoyah Unit 1 Surveillance Capsule Data. .... 14Table 5-3 Reactor Vessel Beltline Material Unirradiated Toughness Properties forSequoyah Unit 1 ............................................................................ ..... 15Table 5-4 Peak Neutron Fluence Projections at Key Azimuthal Locations on the Reactor VesselClad/Base Metal Interface for Sequoyah Unit 1 (x 10'9 n/cm2, B > 1.0 MeV) ............ 16Table 5-5 Sequoyah Unit 1 Calculation of the ART Values for the 1 /4T Location @ 32 EFPY.....17Table 5-6 Sequoyah Unit 1 Calculation of the ART Values for the 3/4T Location @ 32 EFPY ......17Table 5-7 Summary of the Limiting ART Values Used in the Generation of the Sequoyah Unit 1Heatup/Cooldo~vn Curves. .................................................................... 18Table 5-8 RTm-s Calculatiohis for Sequoyah Unit 1 Beitline Region Materials at 32 EFPY.......... 18iii PRESSURE TEMPERATURE LIMITS REPORTList of FiguresFigure 2-1 Sequoyah Unit 1 Reactor Coolant System Heatup Limitations(Heatup Rate of 1 00°F/hr) Applicable for the First 32 EFPY(w/ Margins for Instrumentation Errors of 1 0°F and 60 psig)................................ 4Figure 2-2 Sequoyah Unit 1 Reactor Coolant System Cooldown Limitations(Cooldown Rates up to 100°F/hr) Applicable for the First 32 EFPY(w/ Margins for Instrumentation Errors of 1 0°F and 60 psig)................................ 5Figure 3-1 Sequoyah Unit 1 Selected LTOPS Setpoints ................................................. 11iv PRESSURE TEMPERATURE LIMITS REPORT1.0 RCS Pressure Temperature Limits Report (PTLR)This PTLR for Sequoyah Unit 1 has been prepared in accordance with the requirements of TecirnicalSpecification (TS) 5.6.4. Revisions to the PTLR shall be provided to the NRC after issuance.This report affects TS 3.4.3, RCS Pressure/Temperature Limits (PiT) Limits and TS 3.4.12, LowTemperature Over Pressure Protection (LTOP) System.2.0 RCS Pressure and Temperature LimitsThe limits for TS 3.4.3 are presented in the subsections which follow and were developed using the NRCapproved methodologies specified in TS 5.6.4 with exception of ASME Code Case N-640[13] (Use of Kit),WCAP- 15984-P['14 (Elimination of the Flange Requirement), 1996 Version of Appendix Gm and the-revised fluencesE7T. The operability requirements associated with LTOPS are specified in TS 3.4.12 andwere determined to adequately protect the RCS against brittle fracture in the event of an LTOP Transientin accordance with the methodology specified in TS 5.6.4.2.1 RCS Pressure/Temperature (PIT) Limits (TS 3.4.3)2.1.1 The minimum boltup temperature is 500F2.1.2 The RCS temperature rate-of-change limits are:a. A maximum heatup rate of 1 00°F in any one hour period.* b. A maximum cooldown rate of l00OF in any one hour period.c. A maximum temperature change of less than or equal to I10°F in any one hour period duringinservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.2.1.3 The RCS P/T limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticalityare specified by Figures 2-1 and 2-2.3.0 Low Temperature Overpressure Protection System (TS 3.4.12)The lift setpoints for the pressurizer Power Operated Relief Valves (P.ORVs) are presented in thesubsections which follow. These lift setpoinlts have been developed using the NRC-approvedmethodologies specified in TS 5.6.4.1 PRESSURE TEMPERATURE LIMITS REPORT3.1 Pressurizer PORV Lift Setting LimitsThe pressurizer PORV lift setpoints are specified by Figure 3-1 and Table 3-1 (Ref. 10). The limits for theLTOPS setpoints are contained in the 32 EFPY steady-state curves (Table 2-2), which are beitlineconditions and are not compensated for pressure differences between the pressurizer transmitter and thereactor midplane/beltline or for instrument inaccuracies. The pressure difference between the pressurizertransmitter and the reactor vessel midplane/beltline with four reactor coolant pumps in operation is 68.3psi (Ref. 11).Note: These setpoints include allowance for the pressure difference between the pressurizertransmitter and the reactor vessel midplane/beltline and the SOTF thermal transport effect forheat injection transients. A demonstrated accuracy calculation (Reference 12) has beenperformed to confirn that the setpoints will maintain the system pressure within theestablished limits when the pressure difference between the pressure transmitter and reactormidplane and maximum temperature/pressure instrument uncertainties are applied to thesetpoints.3.2 Anning TemperatureThe LTOPS arming temperature is based upon the methodology defined in the Sequoyah Nuclear PlantUnit 1 Technical Specifications Administrative Controls Section 5.6.4. The arming temperature shall be _<350°F.4.0 Reactor Vessel Material Surveillance ProgramThe reactor vessel material irradiation surveillance specimens shall be removed and examined todetermine changes in material properties. The removal schedule is provided in Table 4-1. The results ofthese examinations shall be used to update Figures 2-1, 2-2 and 3-1.The pressure vessel steel surveillance program (WCAP-8233 r1l) is in compliance with Appendix H to 10CFR 50, "Reactor Vessel Material Surveillance Program Requirementsr21."' The material test requirementsand the acceptance standard utilize the reference nil-ductility temperature which is determined inaccordance with ASTM E23 [3]. The empirical relationship between RTNDr and the fracture toughness ofthe reactor vessel steel is developed in accordance with Code Case N-640 of Section XI of the ASMEBoiler and Pressure Vessel Code, Appendix 0, "Fracture Toughness Criteria for Protection AgainstFailureE41. The surveillance capsule removal schedule meets the requirements of ASTM E185-82N5.Theremoval schedule is provided in Table 4-1.2 PRESSURE TEMPERATURE LIMITS REPORT5.0 Supplemental Data TablesTable 5-1 contains a comparison of measured surveillance material 30 ft-lb transition temperature shiftsand upper shelf energy decreases with Regulatory Guide 1.99, Revision 2[6], predictions.Table 5-2 shows calculations of the surveillance material chemistry factors using surveillance capsuledata. Note that in the calcuilation of the surveillance weld chemistry factor, the ratio procedure fromRegulatory Guide 1.99, Revision 2 was followed. The ratio in question is equal to 0.90.Table 5-3 provides the required Sequoyah Unit 1 reactor vessel toughness data.Table 5-4 provides a summary of the fluence values used in the generation of the heatup and cooldownlimit curves and the PTS evaluation.Table 5-5 and 5-6 show the calculation of the 1/4T and 3/4T adjusted reference temperature at 32 EFPYfor each beitline material in the Sequoyah Unit 1 reactor vessel. The limiting beitline material was theLower Shell Forging 04.Table 5-7 provides a summary of the adjusted reference temperature (ART) Values of the Sequoyah Unit 1reactor vessel beltline materials at the 1/4T and 3/4T locations for 32 EFPY.Table 5-8Sprovides RTPTs values for Sequoyah Unit 1 at 32 EFPY.3 PRESSURE TEMPERATURE LIMITS REPORTMATERIAL PROPERTY BASISLTMITING MATERIAL: LOWER SHELL FORGING 04LIMITING ART VALUES AT 32 EFPY: 1/4T, 216°F3/4T, 186°F2500 --_o- erlim Version.5.1 Run.15680 [i /-2250i Leak Test Limit /-; -2000 .......Unacceptable ....... ... -- --"Acceptable ...,Operation, Operation__:Heatup Rate Critical Limit100o Deg. F/HrI 100 Deg. F/ HrI-- "~2=i(n 'i.120 ,.. .. ....S 1000 --- _ __- ___ __ _750 -- -- -__ _ _ -__ ___-__ __ __ __ --- __/ Criticality Limit based on Iinservice hydrostatic test500 -temnperatuire (2880F) for the __service period up to 32 EFPY(IMinimum !250 ......-~ Boltup .. ... -; -___ __) _STemp = 50*F0 50 100 150 200 250 300 350 400 450 500 550Moderator Temperature (Deg. F)Figure 2-1 Sequoyah Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of100°F/hr)-Applicable for the First 32 EFPY (w/Margins for Instrumentation Errorof 10°F and 60 psig) (Plotted Data provided on Table 2-1)4 PRESSURE TEMPERATURE LIMITS REPORTMATERIAL PROPERTY BASISLIMITING MATERIAL: LOWER SHELL FORGING 04LIMITING ART VALUES AT 32 EFPY: 114T, 216°F3/4T, 186°F2500225020001750150021250o 10007505002500 50 100 150 200 250300 350 400 450 500 550Moderator Temperature (Deg. F)Figure 2-2 Sequoyah Unit 1 Reactor Coolant System Cooldown Limitations (CooldownRates up to 100°F/hr) Applicable for the First 32 EFPY (w/Margins forInstrumentation Error of 10°F and 60 psig) (Plotted Data provided on Table 2-2)5 PRESSURE TEMPERATURE LIMITS REPORTTable 2-1Sequoyah Unit 1 Heatup Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 10°F and 60 psig)100 Heatup 1100 Critica] Limit ILeak Test LimitT P J T Pj T P50 0 288 0 272 200050 477 *288 477 288 248555 477 288 47760 477 288 47765 477 288 47770 477 288 47875 477 288 47880 477 288 48085 477 288 48190 477 ,288 48395 477. 288 485100 477 288 487105 477 288 490110 477 288 493115 477 288 497120 477 288 500125 477 288 505"130 477 288 508135 477 288 515140 477 288 517145 .477 288 527150 477 288 528155 478 288 541160 480 288 541165 483 288 555170 487 288 557175 493 288 571180 500 288 575185 508 288 589190 517 288 609195 528 288 631200 541 288 .656205 555 ,288 684210 571 288 714215 589 .288 748220 609 290 786225 631 295 8286 PRESSURE TEMPERATURE LIMITS REPORTTable 2-1 -(Continued)Sequoyah Unit 1 Heatup Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 1 0°F and 60 psig)100 Heatup 100 Critical LimitT P T P230 656 300 874235 684 305 925240 714 310 981245 748 315 1044250 786 320 1112255 828 325 1188260 874 330 1272265 925 335 1364270 981 340 1466275 1044 345 1578280 1112 350 1702285 1188 355 1838290 1272 360 1988295 1364 365 2154300 1466 370 2337305 1578310 1702315 1838320 1988325 2154330 23377 PRESSURE TEMPERATURE LIMITS REPORTTable 2-2Sequoyah Unit 1 Cooldown Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 10°F and 60 psig)Steady State [20F )40F P1 60F 100F50505560657075808590951001051101151201251301351401451501551601651701751801851901952002052102150552553555556558560561564566569571575578582586591596602608616623632642652664677691707724743764788814843505055606570758085,909510010511011512012513013514014515015516016517017518018519019520020521021505035055075095105125145165185215245275315355405455505565635715795885996106236376526696887097337597878195050556065707580859095100105110115120125130135140145150155160165170175180185190195200205210215045745845946046246446546847047347647948348749249750351051752553454455656858259761463365467770273176279750505560657075808590951001051101151201251301351401451501551601651701751801851901952002052102150408409410411412414416418420423426430434.4384434494564634714794895005125265415585775976206466747057407795050556065707580859095100105110115120125130'1351401451501551601651701751801851901952002052102150305306307308309311.313315318.321325329333338344351358367376"387399412427443461482505530558590624663706754______________________________________ +/- _______________________________________ _________________________________________ I8 PRESSURE TEMPERATURE LIMITS REPORTTable 2-2 -(Continued)Sequoy~a Unit 1 Cooldown Limits at 32 EFPY(without Uncertainties for Instrumentation Errors)Steady State 20F 40F 60F 100F.T P T P T PT P T P220 874 220 853 220 836 220 821 220 806225 909 225 892 225 878 225 869 2-25 865230 948 230 935 230 925 230 921235 991 235 982 235 978240 1038 240 1034245 1090250 1148255 1212260 1283265 1360270 1447275 1542280 1647285 1763290 1892295 2034300 2191305 23649 PRESSURE TEMPERATURE LIMITS REPORT*Table 3-1Selected Setpoints, Sequoyah Unit 1Trs Dg.) PORV#2 PORV#1Trcs(De.F)Setpoint (psig) Setpoint (psig)50 490 465100 500 475135 540 510175 575 540200 610 570250 745 685280 745 685405 745 685450 2350 235010 PRESSURE TEMPERATURE LIMITS REPORTSequoyah Unit I LTOPS Selected Setpoints250020002i1500100005000 50 100 150 200 250 300 350 400 450 500Reactor Coolant System Temnperature (°F)--4-- FOFRV#2 Setpoint --=- PORV#1 SetpointFigure 3-1 Sequoyah Unit 1 Selected LTOPS Setpoints (Plotted Data provided on Table 3-1)11 PRESSURE TEMPERATURE LIMITS REPORTTable 4-1Seqtuoyah Unit 1 Reactor Vessel Surveillance Capsule Withdrawal Schedule* Removal Time FluenceCapsule Location Lead Factorca) (EFPY) (b) (n/cm2,E>l.0 1T 400 3.39 1.03 2.61 x 10O8 (c)U 1400 3.47 3.00 7.96 x 1018 (c)X 2200 3.47 5.27 1.32 x 1019 (c)Y 320° 3.43 10.03 2.19 x 10'9 (c,d)5 40 1.08 Standby (d,e)V 176° 1.08 Standby (d,e)W 184° 1.08 Standby (d,e)Z 3560 1.08 Standby (d,e)Notes:(a)(b)(c).(d)(e)Updated in Capsule Y dosimetry analysis Effective Full Power Years (EFPY) from plant startup.Plant specific evaluation.This fluence is not less than once or greater than twice the peak end of license (32 EFPY) fluenceCapsules 5, V, W and Z will reach a fluence of 2.74 x 1019 (E > 1.0 Mev), the 48 EFPY peakvessel fluence at approximately 44 EFPY, respectively.Administrative Note -The surveillance capsule withdrawal schedule in Table 4-1 is based on thesurveillance program for the original 40 year service life. Relocation of select standby capsules toincrease the fluence lead factor in anticipation of the updated surveillance program for the 60 year licenserenewal service life is described in a TVA letter to NRC dated May 14, 2015 (ML15!34A377).Regulatory approval for the anticipatorY standby capsule relocation has been granted by a NRC letter toTVA dated September 4, 2015 (ML 15244B222). A complete update of the reactor vessel surveillanceprogram for the 60 year license renewal service life will be documented by a subsequent revision to thePTLR prior to entry into the license renewal extended operating period.12 PRESSURE TEMPERATURE LIMITS REPORTTable 5-1Comparison of the Sequoyah Unit 1 Surveillance Material 3 0 ft-lb Transition Temperature Shifts andUpper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions30 ft-lb Transition Upper Shelf EnergyTemperature Shift DecreaseMaterial Capsule Fluence Predicted Measured Predicted Measured____________ ______ (X 1019 n/cm2). (OF)(a) (OF)(b) (%)(a) .(%)(C)Lower Shell T 0.261 .59.85 67.52 16 16Forging 04 U 0.796 89.3 109.7 20.5 21(Tneta)X 1.32 102.6 145.12 23 8(Heat # 980919 /281587) Y 2.19 114.95 129.87 26.5 23Lower Shell T 0.261 59.85 50.59 16 0Forging 04 U 0.796 89.3 67.59 20.5 19(xa)X 1.32 102.6 103.34 23 22(Heat # 980919 / _____281587) Y 2.19 114.95 133.35 26.5 19Weld Metal T 0.261 111.13 127.79 .35 30(Heat # 25295)(d) U 0.796 165.82 144.92 42 26X 1.32 190.51 159.02 .45 .21Y 2.19 213.44 163.8 48 28IiAZ Metal T 0.261 --45.48 --20U 0.796 --. 78.94 --26X 1.32 --95.89 --3Y 2.19 --73.3 --10Notes:(a) Based on Regulatory Guide 1.99, Revision 2, methodology using the mean weight percent values ofcopper and nickel of the surveillance material.(b) .Calculated using measured Charpy data plotted using CVGRAPH, Version 4.1 [8(c) Values are based on the definition of upper shelf energy given in ASTM E185-82.(d) Surveillance Weld was fabricated from weld wire type SMIT 40, Heat # 25295, Flux type SMIT 89,Lot # 1103.13 PRESSURE TEMPERATURE LIMITS REPORTTable 5-2Calculation of Chemistry Factors using Sequoyah Unit 1 Surveillance Capsule DataMaterial Capsule Capsule f~a) FF) ARTNBT(C) FF*ARTNDT FF2Lower Shell T 2.61E+18 0.63 67.52°F 42.54°F 0.40Forging 04 U 7.96E+18 0.94 109.7°F 103.12°F 0.88(Tangential)X 1.32E+19 1.08. 145.12°F 156.73°F 1.16(Heat # 980919/ /_________281587) Y 2.19E+19 1.21 129.87°F 157.14°F 1.47Lower Shell T 2.6 1E+I18 0.63 50.59°F 31.87°F 0.40Forging 04 U 7.96E+18 0.94 67.59°F 63.53°F 0.88(Axial) X 1.32E+19 1.08 103.34°F 111.61°F 1.16(Heat #980919 / *Y 2.19E+19 1.21 133.35°F 161.35°F 1.47281587)SUM: 827.89°F 7.82CF04 =

  • RTNDTr) + 2( FF2) = (827.89) +(7.82) = 105.9°FSurveillance Weld T 2.61E+18 0.63 115.0°F 72.5°F 0.40Material(d) U 7,96E+18 0.94 130.4°F 122.6°F 0.88(Heat # 25295)(e) X 1.32E+19 1.08 143.1°F 154.5°F 1.16Y 2.19E+19 1.21 147.4OF 178.40F 1.47SUM: 528.0°F 3.91CF Surv. we~d =XY(FF
  • RTrrDT) +X( FF2) =(528.0°F) + (3.91) = 135.0°FNotes:(a) f= Calculated fluence from Capsule Y dosimetry analysis resultsETI, (n/cm2, E > 1.0 MeV).(b) FF = fluence factor =fI0.28"0.1logf.(c) ARTNDTValues are the measured 30 ft-lb shift values taken from App. B of Ref. 7, rounded to onedecimal point.(d) The surveillance weld metal ARTNDoT values have been adjusted by a ratio factor of 0.90.(e) Surveillance Weld was fabricated from weld wire type SMIT 40, Heat # 25295, Flux type SMIT 89,Lotft 110314 PRESSURE TEMPERATURE LIMITS REPORTTable 5-3Reactor Vessel Beltline Material Unirradiated Toughness Properties for Sequoyah Unit 1Material Description Cu (%) Ni (%) Initial RTNDT(a)Intermediate Shell Forging 05(Heat'#980807/281489) 01 .640Lower Shell Forging 04(Heat #980919/281587) 01 .670Surveillance Weld (Heat # 25295)(b'd, e) == 0.387 0.11 ---Rotterdam Test(c. e) 0.30 ......-Rotterdam Test(c' e) 0.25 ......Rotterdam Test(c' e) 0.46 ......-Best Estimate of the Intermediate to Lower ShellForging Circumferential Weld Seam W05 0.35 0.11 -40OF(Heat # 25295)(d. e)(a) The Initial RTNDT values are measured values(b) These copper and nickel values are best estimate values for only the surveillance weld metal and is the averageof three data points [0.424 (WCAP-10340, Rev.1), 0.406 (WCAP-10340, Rev.1), 0.33 (WCAP-8233) copperand 0.084 (WCAP-10340, Rev.1), 0.085 (WCAP-10340, Rev.1), 0.17 (WCAP-8233) nickel.]. These values aretreated as one data point in the calculation of the best estimate average for the inter, to lower shell circ. weldshown above. Originally the 0.424 / 0.406 and 0.084 / 0.085 values were reported as single points, 0.41 -0.42and 0.08 (Per WCAP-10340, Rev. l[7d]), but it is actually made up of two data points. Sample TW58 fromCapsule T was broken into two samples, TW58a and TW58b, thus providing the two data points.(c) From NRC Reactor Vessel Integrity Database (RVID) and ultimately fr'om Rotterdam Weld Certifications.(d) Circumferential Weld Seam W05 was fabricated with weld wire type SMvIT 40, Heat # 25295, Flux type SMVIT89, Lot # 2275. The surveillance weld was fabricated with weld Wire type SMIT 40, Heat # 25295, Flux typeSMIT 89, Lot # 1103 and is representative of the intermediate to lower shell circumferential weld.(e) The surveillance weld and the three Rotterdam tests are averaged together for the Best Estimate of theIntermediate to Lower Shell Forging Circumferential Weld Seam.15 PRESSURE TEMPERATURE LIMITS REPORTTable 5-4Peak Neutron Fluence Projections at Key Azimuthal Locations on the Reactor Vessel Clad/Base MetalInterface for Sequoyah Unit 1 (x 10 9 n/cm2, E > 1.0 MeV)Azimuthal LocationEFPY 00 150 300 45010.03 0.205 0.321 0.409 0.63720 0.387 0.596 0.761 1.1832 0.605 0.928 1.19 1.8448 0.896 1.37 1.75 2.7216 PRESSURE TEMPERATURE LIMITS REPORTTable 5-5Sequoyah Unit 1 Calculation of the ART Values for the 1/4T Location @ 32 EFPY(a)Material RG 1:99 CF FF ARTnDT(C) Margin(d) ART(e)R2 Method (0F) (°F) (0F) (0F) (0F)Intermediate Shell Forging 05 Position 1.1 115.6 1.029 40 119.0 34 193Position 1.1 95 1.029 73 97.8 34 205Lower Shell Forging 04Position 2.1 105.9 1.029 73 109.0 34(0 216Intermediate to Lower Shell Position 1.1 161.3 1.029 -40 166.0 56 182Circumferential Weld Seam Position 2.1 135.0 1.029 -40 138.9 56(0 155Notes:(a) Neutron fluence value used for all materials is the highest value from Table 5-4 for 32 EFPY.(b) Initial RTNIJT values are measured values.(c) AXRTNDT = CF
  • FF(d) M = 2 *(a'2 +- OA2)1/2(e) ART = Initial RTNDT + ARTNDT + Margin (0F)(f) Data deemed not-credible (See Reference 7a), thus the full GA will be used to determine margin.Table 5-6Sequoyah Unit 1 Calculation of the ART Values for the 3/4T Location @ 32 EFPY(a)Material RG 1.99 CF FF IRTrmT(b) ART~rTC() Margin(d) ART(e)*R2 Method (0F) (0F) (0F) (0F) (0F)Intermediate Shell Forging 05 Position 1.1 115.6 0.747 40 86.4 34 160Loe hl ogn 4 Position 1.1 *95 0.747' 73 71.0 34 178*Position 2.1 105.9 0.747 73 79.1
  • 34(0 186Intermediate to Lower Shell Position 1.1 161.3 0.747 -40 120.5 56 137Circumferential Weld Seam Position 2.1
  • 135.0 0.747 -40 100.8 56(0 117Notes:(a) Neutron fluence value used for all materials is the highest value from Table 5-4 for 32 EFPY.(b) Initial RTNDT values are measured values.(c) ARTrDT =CF *FF(d) M = 2 *(a.2 + Gra2)1t2(e) ART = Initial RTNDT .+- ARTrNDT + Margin (0F)(f) Data deemed not-credible (See Reference 7a), thus the full GrA will be used to determine margin.17 PRESSURE TEMPERATURE LIMITS REPORTTable 5-7Summnary of the Sequoyah Unit 1 Reactor Vessel Beitline Material ART ValuesMaterial RG 1.99 R2 1/4 ART 3/4 ARTMethod (0F) (0F)Intermediate Shell Forging 05 Position 1.1 193 160Position 1.1 205 178Lower Shell Forging 04Position 2.1 216 186Intermediate to Lower Shell Position 1.1 182 137Circumferential Weld Seam Position 2.1 155 117Table 5-8RTP-s Calculations for Sequoyah Unit 1 Beltline .Region Materials at 32 EFPY(a)Material Fluence FF CF ARTpTs~b) Margin RTpxs(d)(X 1019 n/cm2, "(F) (0F) (0F) (OF) (0F)E>1.0 MeV)Intermediate Shell Forging 05 1.84 1.167 115.6 .134.9 34 40 209Lower Shell Forging 04 1.84 1.167 95.0 110.9 34 73 218Lower Shell Forging 04 1:84 1.167 105.9 123.6 34(e) 73 231(Using S/C Data)Circumferential Weld Metal 1.84 1.167 161.3 188:2 56 -40 204Circumferential Weld Metal 1.84 1.167 135.0 157.5 56(e) -40 174(Using S/C Data)Notes:(a) Neutron fluence value used for all materials is the highest value from Table 5-4 for 32 EFPY.(b) ARTPTs = CF *FF(c) Initial RTNOT values are measured values(d) RTPTs= + ARTpTs + Margin (0F)(e) Data deemed not-credible (See Reference 7a), thus the full will be used to determine margin.18 PRESSURE TEMPERATURE LIMITS REPORT6.0 References1. WCAP-8233, Tennessee Valley Authority Sequoyah Unit No. 1 Reactor Vessel RadiationSurveillance Progr~am, S. E. Yanichko, et. al., December 1973.2. Code of Federal Regulations, 10OCFR50, Appendix H, Reactor Vessel Material SurveillancePro gramn Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C..3. ASTM E23 Standard Test Method Notched Bar himpact Testing of Metallic Materials, in ASTMStandards, American Society for Testing and Materials, Philadelphia, PA.4.Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G, Fracture Toughness"Criteria for Protection Against Failure5. ASTM E185-82, Annual Book of ASTM Standards, Section 12, Volume 12.02, Standard Practicefor Conducting Surveillance Tests for Light- Water Cooled Nuclear Power Reactor Vessels.6. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S.Nuclear Regulatory Commission, May 1988.7a. WCAP-15224, Analysis of Capsule the Tennessee Valley Author"ity Sequoyah Unit 1Reactor Vessel Radiation Surveillance Program, T.J. Laubham, et. al., June 1999.7b. WCAP-1 3333, Analysis of Capsule Xfoino the Tennessee Valley Authority Sequoyah Unit 1Reactor Vessel Radiation Surveillance Progr~am, M.A. Ramirez, S. L. Anderson, L. Albertin, June1992.7c. .SwRI Project 06-8851, Reactor Vessel Material Surveillance Progr'am for Sequoyah Unit No. 1:Analysis of Capsule U, P. K. Nair, et al., October 1986.7d. WCAP- 10340, Revision 1, Analysis of Capsule T foin the Tennessee Valley Authority SequoyahUnit.] Reactor Vessel Radiation Surveillance Program, S.E. Yanichko, et. al., February 1984.8. CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.1, developed by ATIConsulting, March 1999.9. WCAP-14040-NP-A, Revision 2, Methodology Used to Develop Cold Overpressure MitigatingSystem Setpoints and RCS Heatup and Cooldown Limit Curves, J.D. Andrachek, et. al., January1996.10. WCAP-15293, Revision 2, Sequoyah Unit 1 Heatup and Cooldown Limit Curves for NormalOperation and PTLR Support Documentation, J.H. Ledger, July 2003.11. Westinghouse Letter to TVA, TVA-93 -105, Cold Overpressure Mitigation System Code Case andDelta-P Calculation, dated May 19, 1993.12. Calculation SQN-IC-01 4, Demonstrated Accuracy Calculation for Cold Overpressure ProtectionSystem.19 PRESSURE TEMPERATURE LIMITS REPORT13. ASME Code Case N-640, Alternative Reference Fracture Toughness for Development of P-TLimit Curves for Section XL, Division 1, dated February 26, 1999.14. WCAP-15984-P, Revision 01, Reactor Vessel Closure Head/Vessel Flange RequirementsEvaluation for Sequoyah Units 1 and 2, W. Bamford, et.al., April 2003.20 ENCLOSURE 3SEQUOYAH UNIT 2PRESSURE TEMPERATURE LIMITS REPORT, REVISION 6 Bi88 151.016"PRESSURE TEMPERATURE LIMITs REPORT801IContractrfo ny port of his re-.dligdealsad dinmsnsions. .*: 'Octobe~r 16,.2015. _* SOEP (Nt) BYW. J.Pierce__*Tennessee Valley AuthoritySequoyah unit 2Pressure Temperature Limits ReportRevision 6, september 2015PROJECT Secjuovah 'DisCIPLINE NCONTRACT " 4411 UNIT. 2DESC. RCS Pressure-Temperature L'mait-ReportDWG/DOC NO. PTLR-2SHEET ..- OF -REV. 06*DATE. 10/16/15 ECN/DCN FILE N2N-081"EDMS, WT CA-K PRESSURE TEMPERATURE LIMITS REPORTTable of ContentsLito T b e .List. .................of.............Tables........... .......................iList of Figures ............................................................................................. ........ v1.0 RCS Pressure Temperature Limits RePort (PTLR) ........ .... .................................... 12.0 Operating Limits ............................................................................. .............12.1 RCS Pressure/Temperature (PIT) Limits (TS 3.4.3)................................................. 13.0 Low Temperature Overpressure Protection System (TS 3.4.12) ...................... ..13.1 Pressurizer PORV Lift Setting Limits ........... .....................3.2. Arming Temperature ............'......24.0 Reactor Vessel Material Surveillance Program.....,................................................. 25.0 Supplemental Data Tables ............................................................................3..36.0 References ......................................... .. ... ..... ........ 17ii PRESSURE TEMPERATURE LIMITS REPORTList of TablesTable 2-1 Sequoyah Unit 2 Heatup Limits at 32 BEFPY(with Uncertainties for Instrumentation Errors of l0°F and 60 psig)............................... 6Table 2-2 Sequoyah Unit 2 Cooldown Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 10°F and 60 psig)...... ......................... 7Table 3.-1 Selected Setpoints, Sequoyah Unit 2........................................................... 8Table 4-1 Sequoyah Unit 2 Reactor Vessel Surveillance Capsule Withdrawal Schedule............. 10Table 5-1 Comparison of the Sequoyah Unit 2 Surveillance Material 30 ft-lb TransitionTemperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99,revision 2, Predictions................ .......................................................... 11Table 5-2 Calculation of Chemistry Factors using Sequoyah Unit 2 Surveillance Capsule Data....12Table 5-3 Reactor Vessel Beltline Material Unirradiated Toughness Properties forSequoyah Unit2 ................................................................................ 13Table 5-4 Peak Neutron Pluence Projections at Key Azimuthal Locations on the Reactor VesselClad/Base Metal Interface for Sequoyah Unit 2 (x 1019 n/cm2, EB> 1.0 MeV) ....i........ 14Table 5-5 Sequoyah Unit 2 Calculation of the ART Values for the 1/4T Location @ 32 EFPY.....15Table 5-6 Sequoyah Unit 2 Calculation of the ART Values for the 3/4T Location @ 32 EPPY.....15Table 5-7 Summary of the Limiting ART Values Used in the Generation of the Sequoyah Unit 2Heatup/Cooldown Curves.......................................................... :............ 16Table 5-8 RTPTs Calculations for Sequoyah Unit 2 Beltline Region Materials at 32 EFPY.......... 16iii PRESSURE TEMPERATURE LIMITS REPORTList of FiguresFigure 2-1 Sequoyah Unit 2 Reactor Coolant System Heatup Limitations(Heatup Rate of I 00°F/hr) Applicable for the First 32 EFPY(wi/Margins for'Instrumentation Errors of 1 00F and 60 psig)................................ 4Figure 2-2 Sequoyah Unit 2 Reactor Coolant System Cooldown Limitations(Cooldown Rates up to 1 00°F/hr) Applicable for the First 32 EFPY(w/ Margins for Instrumentation Errors of 10°F and 60 psig) .................... .............5Figure 3-1 Sequoyah Unit 2 Selected LTOPS Setpoints .................................................. 9iv PRESSURE TEMPERATURE LIMITS REPORT1.0 RCS Pressure Temperature Limits Report (PTLR)This PTLR for Sequoyah Unit 2 has been prepared in accordance with the requirements of TechnicalSpecification (TS) 5.6.4. Revisions to the PTLR shall be provided to the NRC after issuance.This report affects TS.3.4.3, RCS Pressure/Temperature Limits (P/T) Limits and TS 3.4.12,.LowTemperature Overpressure Protection (LTOP) System.2.0 RCS Pressure and Temperature LimitsThe limits for TS 3.4.3 are presented in the subsections which follow and were developed using the NRCapproved methodologies specified in TS 5.6.4 with exception of ASME Code Case N-64011n1 (Use of K1o),WCAP- 159 84-Ptl21 (Elimination of the Flange Requirement), 1996 Version of Appendix G[41 and therevised fluencesETI. The operability requirements associated with LTOPS are specified in TS 3.4.12 andwere determined to adequately protect the RCS against brittle fracture in the event of an LTOP Transientin accordance with the methodology specified in TS 5.6.4.2.1 RCS Pressure/Temperature (P/T) Limits (TS 3.4.3)2.1.1 The minimum boltup temperature is 50°F2.1.2 The RCS temperature rate-of-change limits are:a. A maximum heatup rate of 1 00°F in any one hour period.b. A maximum cooldown rate of 100°F in any one hour period.c. A maximum temperature change of less than or equal to. 1 0°F in any one hour period duringinservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.2.1.3 The RCS P/T limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticalityare specified by Figures 2-1 and 2-2.3.0 Low Temperature Overpressure Protection System (TS 3.4.12)The lift setpoints for the pressurizer Power Operated Relief Valves (PORVs) are presented in thesubsections which follow. These lift setpoints have been developed using the NRC-approvedmethodologies specified in TS 5.6.4.I PRESSURE TEMPERATURE LIMITS REPORT*3.1 Pressurizer PORV Lift Setting LimitsThe pressurizer PORV lift setpoints are specified by Figure 3-1 and Table 3-1 (Ref. 10). The limits for theLTOPS setpoints are contained in the 32 EFPY steady-state curves (Table 2-2), which are beitlineconditions and are not compensated for pressure differences between the pressurizer transmitter and thereactor midplane/beltline or for instrument inaccuracies. The pressure difference between the pressurizertransmitter and the reactor vessel midplane/beltline with four reactor coolant pumps in operation is 68.3psi (Ref. 13).Note: These set-points include allowance for the pressure difference between the pressurizertransmitter and the reactor vessel midplane/beltline and the 50°F thermal transport effect forheat injection transients.. A demonstrated accuracy calculation (Reference 14) has beenperformed to confirm that the setpoints will maintain the system pressure within theestablished limits when the pressure difference between the pressure transmhitter and reactormidplane and maximum temperature/pressure instrument uncertainties are applied to thesetpoints.3.2 Arming TemperatureThe LTOPS arning temperature is based upon the methodology defined in the Sequoyah Nuclear PlantUnit 2 Technical Specifications Administrative Controls Section 5.6.4. The arming temperature shall be <350°F.4.0 Reactor Vessel Material Surveillance ProgramThe reactor vessel material irradiation surveillance specimens shall be removed and examined todetermine changes in material properties. The removal schedule is provided in Table 4-1. The results ofthese examinations shall be used to update Figures 2-1, 2-2 and 3-1.The pressure vessel steel surveillance program (WCAP-85 13[1]) is in compliance with Appendix H to 10CFR 50, "Reactor Vessel Material Surveillance Program Requirementst21."' The material test requirementsand the acceptance standard utilize the reference nil-ductility temperature RTNDT, which is determined inaccordance with ASTM E23 [3. The empirical relationship between RTNOT and the fracture toughness ofthe reactor vessel steel is developed in accordance with Code Case N-640 of Section XI of the ASMEBoiler and Pressure Vessel Code, Appendix G, "Fracture Toughness Criteria for Protection AgainstFailurer4k.' The surveillance capsule removal schedule meets the requirements of ASTM E185-82N5.Theremoval schedule is provided in Table 4-1.2 PRESSURE TEMPERATURE LIMITS REPORT5.0 Supplemental Data TablesTable 5-1 contains a comparison of measured surveillance material 30 ft-lb transition temperature shiftsand upper shelf energy decreases with Regulatory Guide 1.99, Revision 2[6], predictions.Table 5-2 shaows calculations of the surveillance material chemistry factors using surveillance capsuledata. Note that in the calculation of the surveillance weld chemistry factor, the ratio procedure fromRegulatory Guide 1.99, Revision 2 was followed. The ratio in question is equal to 0.93.Table 5-3 provides the required Sequoyah Unit 2 reactor vessel toughness data.Table 5-4 provides a summary of the fluence values used in the generation of the heatup and cooldownlimit curves and the PTS evaluation.Table 5-5 and 5-6 show the calculation of the 1/4T and 3/4T adjusted reference temperature at 32 EFPYfor each beltline material in the Sequoyah Unit 2 reactor vessel. The limiting beltline material was theIntermediate Shell Forging 05..Table 5-7 provides a summary of the adjusted reference temperature (ART) values of the Sequoyah Unit 2reactor vessel beltline materials at the l/4T and 3/4T locations for 32 EFPY.Table 5-8 provides RTP-Ts values for Sequoyah Unit 2 at 32 EFPY.3 PRESSURE .TEMPERATURE LIMITS REPORTMATERIAL PROPERTY BASISLIMITING MATERIAL! INTERMEDIATE SHELL FORGING 05LIMITING ART VALUES AT 32 EFPY: 1/4T, 142°F3/4T, l15°F2Rflfl-v Io)perlim Version:5.1 Run:5694 /ILeak, Test, Limit 22502000--[ Unacceptable1Oeain ....*Acceptable ...Operation1750 -..1500125U) 00Co0 25~ritical Limit*100 Deg. F/Hr750"500250MinumumBoltup Temp-- = 500F "Criticality Limit based on-~inservic e hydrostatic testtemperature (214°F) for theservice period up to 32 EFPYUf.......I ..............! ....! ........ .........0 50 100 150 200 250 300 350 400 450 500 550Moderator Temperature (Deg. F)Figure 2-1 Sequoyah Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of1000F/hr) Applicable for the First 32 EFFY (w/ Margins for InstrumentationError of 10°F and 60 psig) (Plotted Data provided on Table 2-1)

PRESSURE TEMPERATURE LIMITS REPORTMATERIAL PROPERTY BASISLIMITING MATERIAL: INTERMEDIATE SHELL FORGING 05.LIMITING ART VALUES AT 32 EFPY: 1/4T, 142°F3/4T, 150F.2500 [Oeri Version:5.1 Run:5694]2000 tUnacceptable _____ --__ _________ AcceptableOperation OPeration -1750 ..... 'S 1250 * -.-o----__ -__ __ ____ '--"o Cooldown~~Rates i"= 1000 F/Hr .......................u steady-stateo-6o1750 60 oo250 -.... ...........M inim um I- ......:.. ........ .. ... ...4.. ... .........SBoltup Templi ..! ' i-500F !0 50 100 150 200 250 300 350 400 450 500 550Moderator Temperature (Deg. F)Figure 2-2 Sequoyahi~nit 2 Reactor Coolant System Cooldown Limitations (CooldownRates up to 100°F/hr) Applicable for the First 32 EFPY (w/ Margins forInstrumentation Error of 100F and 60 psig) (Plotted Data provided on Table 2-2)2 PRESSURE TEMPERATURE LIMITS REPORTTable 2-1Sequoyah Unit 2 Heatup Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 10°F and 60 psig)100 Heatup 1100 Critical Limit ILeak Test LimitT P T P JT P50 0 214 0 198 200050 591 214 607 214 248555 595 214 61460 601 214 62265 607 214 65770 614 214 65075 622 214 64780 630 214 64685 640 214 64890 646 214 *65395 646 214 661100 6.46 214 671105 646 214 680110 646 214 685115 646 214 701120 646 214 720125 648 214 743130 653 214 769135 661 214 798140 671 215 832145 685 220 869150 701 225 911155 720 230 959160. 743 235 1011165 769 240 1069170 798 245 1134175 832 250 1206180 869 255 1286185 911 260 1374190 959 265 1471195 1011 270 1579200 1069 275 1698205 1134 280 1829210 1206 285 1974215 1286 290 2134220 1374 295 2311225 1471230 1579235 1698240 1829245 1974250 2134255 23113 PRESSURE TEMPERATURE LIMITS REPORTTable 2-2Sequoyah Unit 2 Cooldown Limits at 32 EFPY(with Uncertainties for Instrumentation Errors of 1 0°F and 60 psig)Steady State 1 20F 40F 60F I00OFT P T P T P T P T P5050556065707580859095100105110115120125130135140145*1501551601651761751801851901952002052102152202252300591595601607614622630640650661674688703720739.7607838098378689029409821028108011361199126813441429152216251739186520042158232850505560657075808590951001051101151201251301351401451501551601650552554558564572580589599610623636652668687707730755783814848885927.973102450505560657075808590951001051101151201251301351401451501551600503508514521529538548*55957158459961663465467670172975979383187291896850505560657075808590951001051101151201251301351401451501550.461466470478486496506518531546562580600622647674704738775816862913505055606570758085909510010511011512012513013514014515003663723803893994104234374534704905125365635936266637047498008564 PRESSURE TEMPERATURE LIMITS REPORTTable 3-1Selected Setpoints, Sequoyah Unit 2PORV#2 PORV#1Trcs (Deg.F) Setpoint (psig) Setpoint (psig)50 510 485100 580 555135 640 610174 745 682200 745 685*250 745 *685-278 745 685400 745 685450 2350 23505 PRESSURE TEMPERATURE LIMITS REPORTSequoyah Unit 2 LTOPS Selected Setpoints-/! [4-2500 --....J Lo- I I !0 01015 0 5 303040 5 0Reacto Coln Iyte enpraue(FA F--*-- FORV#2-FinaI --U-- FORV#1Figure 3-1 Sequoyah Unit 2 Selected LTOPS Setpoints (Plotted Data pro vided on Table 3-1)6 PRESSURE TEMPERATURE LIMITS REPORTTable 4-1Sequoyah Unit 2 Reactor Vessel Surveillance Capsule Withdrawal Schedule(a) Updated in Capsule Y dosimetry analysis (WCAP-15320t71).(b) Effective Full Power Years (EFPY) from plant startup.(c) Plant specific evaluation.(d) This fluence is not less than once or greater than twi ce the peak end of license (32 EFPY) fluence(e) Capsules 5, V, W and Z will reach a fluence of 2.71 x 1019 (EB> 1.0 MeV), the 48 EFPY peakvessel fluence at approximately 44 EFPY.,Administrative Note -The surveillance capsule withdrawal schedule in Table 4-1 is based on thesurveillance program for the original 40 year service life. Relocation of select standby, capsules toincrease the fluence lead factor in anticipation of the updated surveillance program for the 60 year licenserenewal service life is described in a TVA letter to NRC dated January 10, 2013 (ML13032A25 1). .Regulatory approval for the anticipatory standby capsule relocation has been granted by a NRC letter toTVA dated September 27, 2013 (ML13240A320). A complete update of the reactor vessel surveillanceprogram for the 60 year license renewal service life will be documented by a subsequent revision to thePTLR prior to entry into the license renewal extended operating period.7 PRESSURE TEMPERATURE LIMITS REPORTTable 5-1Comparison of the Sequoyah Unit 2 Surveillance. Material 30 ft-lb Transition Temperature Shifts andUpper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions30 ft-lb Transition Upper Shelf EnergyTemperature Shift DecreaseMaterial Capsule Fluence ' Predicted Measured Predicted Measured(x 1019 n/cm2) (OF)(a) (OF)(b) (%,/)(al) (%)(c)Intermediate Shell T 0.261 60.33. 63.65 17 12Forging 05 U 0.692 85.22 79.31 21 16(Tangential)(et285/X 1.22 100.23 85.7 23 8981057) Y 2.14 114.67 134.12 26 22Intermediate Shell T 0.261 60.33 48.73 17 7Forging 05 U 0.692 85.22 66.06 21 9(Axial)(et285/X 1.22 100.23 110.04 23 2981057) Y 2.14 .114.67 89.21 26 22Weld Metal T 0.261 43.12 74.56 20 2(Heat # 4278)(d) U 0.692 60.91 130.38 25 6X 1.22 71.63 44.22 29 35Y 2.14 81.96 86.91 33 "3HAZ Metal T 0.261 --24.58 -,- 2U 0.692 --64.03 --14X 1.22 --28.29 --19Y 2.14 --50.32 --39Notes:(a) Based on Regulatory Guide 1.99, Revision 2, methodology using the mean weight percent values ofcopper 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) Surveillance Weld was fabricated from weld wire type SMIT 89, Heat # 4278, Flux type SMIT 89,Lot # 1211.8 PRESSURE TEMPERATURE LIMITS REPORTTable 5-2Calculation of Chemistry Factors using Sequoyah Unit 2 Surveillance Capsule DataMaterial Capsule Capsule f(a) FF(b) ARTNDT(C) FF*ARTNDT FF2Intermediate Shell T 2.61E+18 0.635 63.7 40.45 0.403Forging 05 U 6.92E+18 0.897 79.3 71.13 0.805(Tangential) X 1.22E+19 1.055 85.7 90.41 1.113(Heat #288757 / Y 2.14E+19 1.207 134.1 161.86 1.457981057)Intermediate Shell T 2.61E+18 0.635 48.7 30.92 0.403Forging 05 U 6.92E+18 0.897 66.1 -59.29 0.805(Axial) X 1.22E+19 1.055 110.0 116.05 1.113(Heat #288757 / Y 2.14E+19 1.207 89.2 107.66 1.457981057) ______ ______SUM: 677.77°F 7.556CFo5 = X(FF

  • RTNDT) + X.( FF2) = (677.77) +(7.556) = 89.70FSurveillance Weld T 2.6 1E+18 0.635 69.4 (74.6) 44.07 0.403Material(d) *U 6.92E+18 0.897 121.3 (130.4) 108.81 0.805(Heat # 4278)(e) X 1.22E+19 1.055 41.1 (44.2) 43.36 1.113Y 2.14E+19 1.207 80.8 (86.9) 97.53 1.457SUM: 293 .77°F 3.778CF Surv. weld =
  • RTNDT) + X( FF2) = (293.77°F) +(3.778) = 77.8"FNotes:(a) f = Calculated fluence from Capsule Y dosimetry analysis results [7], (n/cma2, E > 1.0 MeV).(b) FF = fluence factor = f02-.~o 3(c) AIRTNDT values are the measured 30 ft-lb shift values taken from App. B of Ref. 7, rounded to onedecimal point.(d) The surveillance weld metal ARTNDT Values have been adjusted by a ratio factor of 0.93.(e) Surveillance Weld was fabricated from weld wire type SMIT 89, Heat # 4278, Flux type SMIT 89, Lot #1211.

PRESSURE TEMPERATURE LIMITS REPORTTable 5-3Reactor Vessel Beltline Material Unirradiated TouglmessProperties for Sequoyah Unit 2Material Description Cu (%) Ni (%) Initial RTNDT(a)Intermediate Shell Forging 05(Heat #288757 /981057)013.710FLower Shell Forging 040.14 0.76 -22 °F(Heat # 990469 / 293323)Intermediate to Lower Shell ForgingCircumferential Weld Seam W05(b) 0.12 0.11 -40F(Heat # 4278)Surveillance Weld(b) 0.13 0.11Notes:(a) The Initial RTNDT values are measured values(b) Circumferential Weld Seam was fabricated with weld wire type SMIT 89, Heat # 4278, Flux type SMIT89, Lot # 1211 and is representative of the intermediate to lower shell circumferential weld.10 PRESSURE TEMPERATURE LIMITS REPORTTable 5-4Peak Neutron Fluence Projections at Key Azimuthal Locations on the Reactor Vessel Clad/Base MetalInterface for Sequoyah Unit 2 (x 1019 n/cm2, E > 1.0 MeV)Azimuthal LocationEFPY 0° 150 3 00 45010.54 0.211 0.336 0.426 0.63720 0.38 0.60 0.773 1.1632 0.593 0.934 1.21 1.8248 0.878 1.38 1.80 2.7111 PRESSURE TEMPERATURE LIMITS REPORTTable 5-5Sequoyah Unit 2 Calculation of the ART Values for the 1/4T Location @ 32 EFPY(a)Material RG 1.99 cF FF IRTNDT(b) ARTNDTo(C) Margin(d) ART(e)R2 Method (0F) _____ (0F) (0F) (0F) (0F)Position 1.1 95 1.027 10 97.6 34 142Intermediate Shell Forging 05Position 2.1 89.7 1.027 10 92.1 34 136Lower Shell Forging 04 Position 1.1 104 1.027 -22 106.8 34(0 119Intermediate to Lower Shell " Position 1.1 .63 '1.027 -4 64.7 56 117Circumferential Weld Seam Position 2.1 77.8 " 1.027 .-4 79.9 56(0 132Notes:(a) Neutron fluence value used for all materials is the highest value from Table 5-4 for 32 EFPY.(b) Initial RTNDoT values are measured values.(c) ARTrmT CF *FF(d) M =2 *(a'i2 +I 0'A2)1/2(e) ART =Initial RTNDT +- ARTNDTr + Margin (0F)(f) Data deemed not-credible (See Reference 7a), thus the full ca will be used to determine margin.Table 5-6Sequoyah Unit 2 Calculation of the ART Values for the 3/4T Location @ 32 EFPY(a)Material RG 1.99 CF FF IRTNDT(b) Margin(d) ART(e)R2 Method (0F) (0F) (0F) (0F) (0F)Position 1.1 95 0.745 10 70.8 34

  • 115Intermediate Shell Forging 05Position 2.1 89.7 0.745 10 66.8 34 .111Lower Shell Forging 04 Position 1.1 104 0.745 -22 77.5 34(0 90intermediate to Lower Shell Position 1.1 63 0.745 .-4 46.9 56 99Circumferential Weld Seam Position 2.1 77.8 0.745 -4 58.0 56(0 110Notes: .(a) Neutron fluence value used for all materials is the highest value from Table 5-4 for 32 EFPY.(b) Initial RTrNDT values are measured values.(c) =CF *FF(d) M = 2 *(0i2 + raA)2(e) ART =Initial RTrNDv + ARTNDT + Margin (0F)(f) Data deemed not-credible (See Reference 7a), thus the full GA will be used to determine margin.12 PRESSURE TEMPERATURE LIMITS REPORTTable 5-7Summary of the Sequoyah Unit 2 Reactor Vessel Beitline Material ART ValuesMaterial RG 1.99 R2 1/4 ART 3/4 ARTMethod (0F) (0F)Position 1.1 142 115Intermediate Shell Forging 05Position 2.1 136 111Lower Shell Forging 04 Position 1.1 119 90Intermediate to Lower Shell Position 1.1 117 99Circumferential Weld Seam Position 2.1I 132 110Table 5-8RTPTs Calculations for Sequoyah Unit 2 BeltlineRegion Materials at 32 EFPY(a)Material Fluence FF CF ARTpTs~b) Margin RTNDT(U-)(C) RTpTs(d)(x 10 19 n/cm2, (0F) (0F) (0F) (0F) (0F)E>l.0 MeV)Intermediate Shell Forging 05 1.82 1.164 95 110.6 34 10 155Intermediate Shell Forging 05 1.82 1.164 89.7 104.4 34 10 148(Using S/C Data) _______Lower Shell Forging 04 1.82 .1.164 104 121.1 34(e) -22 133Circumferential Weld Metal 1.82 1.164 63 73.3 56 -4 125Circumferential Weld Metal 1.82 1.164 77.8 90.6 56(e) -4 143(Using S/C Data)____Notes:(a) Neutron fluence value used for all materials is the highest value from Table 5-4 for 32 EFPY.(b) ARTpTs= CF
  • FF(c) Initial RTNDT values are measured values(d) RTpTs = RTNDT(U) + ARTpTs + Margin (0F)(e) Data deemed not-credible (See Reference 7a), thus the full ra.. will be used to determine margin.13 PRESSURE TEMPERATURE LIMITS REPORT6.0 References1. WCAP-8513, Tennessee Valley Authority Sequoyah Unit No. 2 Reactor Vessel RadiationSurveillance Program, J. A. Davidson, et. al., November 1975.2. Code of Federal Regulations, 10OCFR50, Appendix H, Reactor Vessel Material SurveillanceProgr~am Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C.3. ASTM E23 Standard Test Method Notched Bar" Impact Testing of Metallic Materials, in ASTMStandards, American Society for Testing and Materials, Philadelphia, PA.4.Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G, Fracture ToughnessCriteria for Protection Against Failure5. ASTM E185-82, Annual Book of ASTM Standards, Section 12, Volume 12.02, Standard Practicefor Conducting Surveillance Tests for Light- Water Cooled Nuclear Power Reactor Vessels.6. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S.Nuclear Regulatory Commission, May 1988.7a. WCAP- 15320, Analysis of Capsule Y fi'omn the Tennessee Valley Authority Sequoyah Unit 2Reactor Vessel Radiation Surveillance Program, T.J. Laubham, et. al., November 1999.7b. WCAP-*10509, Analysis of Capsule T fiom the. Tennessee Valley Authority Sequoyah Unit 2Reactor Vessel Radiation Surveillance Program, R. S. Boggs, et al, April 1984.7c. Southwest Research Institute Nondestructive Evaluation Science and Technology Division,Reactor Vessel Material Surveillance Program and Technology Division, Reactor Vessel Material"Surveillance Pro gram for Sequoyah Unit 2." Analysis of Capsule U, Final Report SwRI Project*17-8851 TVA Contra~ct 85PJH-964430, January !1990.7d. WCAP- 13545, Analysis of Capsule X from the Tennessee Valley Authority Sequoyah Unit 2Reactor Vessel Radiation Surveillance Program, M. A. Ramirez, S. L. Anderson, A. Madeyski,November 1992.8. CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.1, developed by ATIConsulting, March 1999.9. WCAP- 14040-NP-A, Revision 2, Methodology Used to Develop Cold Overpressure MitigatingSystem Setpoints and RCS Heatup and Cooldown Limit Curves, J.D. Andrachek, et. al., January1996.10. WCAP-15321, Revision 2, Sequoyah Unit 2 Heatup and Cooldown Limit Curves for NormalOperation and PlTLR Support Documentation, J.H. Ledger, et.al., July 2003.11. ASME Code Case N-640, Alternative Reference Fracture Toughness for Development of P-T*Limit Curves for Section XL, Division 1, dated February 26, 1999.14 PRESSURE TEMPERATURE LIMITS REPORT12. WCAP-15984-P, Revision 01, Reactor Vessel Closure Head/Vessel Flange RequiremnentsEvaluation For Sequoyah Units 1 and 2, W. Bamford, et.al., April 2003.13. Westinghouse Letter to TVA, TVA-93-105, Cold Overpressure Mitigation System Code Case andDelta-P Calculation, dated May 19, 1993.14. Calculation SQN-IC-0 14, Demonstrated Accuracy Calculation for Cold Overpressure ProtectionSystem.15
Retrieved from "https://kanterella.com/w/index.php?title=ML15322A055&oldid=693546"