ML18352A229

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Pressure Temperature Limits Report, Revision 7
ML18352A229
Person / Time
Site: Sequoyah Tennessee Valley Authority icon.png
Issue date: 12/17/2018
From: Anthony Williams
Tennessee Valley Authority
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML18352A229 (27)
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Text

Tennessee Valley Authority, Sequoyah Nuclear Plant, P.O. Box 2000, Soddy Daisy, Tennessee 37384 December 17, 2018 10CFR50.4 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Sequoyah Nuclear Plant, Unit 2 Renewed Facility Operating License Nos. DPR-79 NRC Docket Nos. 50-328

Subject:

Sequoyah Unit 2 Pressure Temperature Limits Report, Revision 7 In accordance with Sequoyah Nuclear Plant (SQN) Unit 2 Technical Specifications (TSs) 5.6.4.C, enclosed is the Unit 2 Pressure Temperature Limits Report (PTLR), Revision 7. The Unit 2 PTLR has been revised for license renewal actions including: an updated neutron fluence evaluation to extend pressure-temperature (P-T) limits from 32 effective full power year (EFPY) to 44.7 EFPY; evaluation of the reactor vessel materials in the extend beltline region; and revision to the surveillance capsule withdrawal schedule. The low temperature overpressure protection system setpoints have been revised for the pressurizer power operated relief valve (PORV) digital control system upgrade. The PORV digital control system upgrade was implemented on November 21, 2018 during the Unit 2 Cycle 22 Refueling Outage. Associated administrative changes were made for report consistency.

There^HFeHTOH^ewj^gulatory commitments in this letter. If you have anyquestions, please Contact Jonathan JcSfyison, SQN Site Licensing Manager at (423) 843-8129.

Respectfully, Anthony L. Williams Site Vice President Sequoyah Nuclear Plant Enclosure

1. Sequoyah Unit 2 Pressure Temperature Limits Report, Revision 7 printed on recycled paper

ENCLOSURE 1 SEQUOYAH UNIT 2 PRESSURE TEMPERATURE LIMITS REPORT, REVISION 7

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-rfSMesseE <<<<i<<r^uo<<pirir SQEP (N) BY: W.ChrfcBteneau Tennessee Valley Authority Sequoyah Unit 2 Pressure Temperature Limits Report Revision 7, September 2017 PROJECT Sequovah DISCIPLINE N CONTRACT 4411 UNIT DESC. RCS Pressure-Temperature Limit Report DWG/DOC NO. PTLR-2 SHEET - OF REV. 07 DATE 04/30/18 ECN/DCN FILE N2N-081 EDMS, WT CA-K

PRESSURE TEMPERATURE LIMITS REPORT Table of Contents List of Tables

.IV List of Figures v 1.0 RCS Pressure Temperature Limits Report (PTLR) 1 2.0 Operating Limits j 2.1 RCS Pressure/Temperature (P/T) Limits (TS 3.4.3) 1 3.0 Low Temperature Overpressure Protection System (TS 3.4.12) 2 3.1 Pressurizer PORV Lift Setting Limits 2 3.2 Arming Temperature 2 4.0 Reactor Vessel Material Surveillance Program 2 5.0 Supplemental Data Tables 3 6.0 References 20

PRESSURE TEMPERATURE LIMITS REPORT List of Tables Table 2-1 Sequoyah Unit 2 Heatup Limits at 44.7 EFPY (with Uncertainties for Instrumentation Errors of 10°F and 60 psig) 6 Table 2-2 Sequoyah Unit 2 Cooldown Limits at 44.7 EFPY (with Uncertainties for Instrumentation Errors of 10°F and 60 psig) 7 Table 3-1 Selected Setpoints, Sequoyah Unit 2 8 Table 4-1 Sequoyah Unit 2 Reactor Vessel Surveillance Capsule Withdrawal Schedule 10 Table 5-1 Comparison of the Sequoyah Unit 2 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions 11 Table 5-2 Calculation of Chemistry Factors using Sequoyah Unit 2 Surveillance Capsule Data 12 Table 5-3 Reactor Vessel Beltline and Extended Beltline Material Unirradiated Toughness Properties for Sequoyah Unit 2 13 Table 5-4 Peak Neutron Fluence Projections on the Reactor Vessel Clad/Base Metal Interface for Sequoyah Unit 2 (x 1019 n/cm2, E > 1.0 MeV) 14 Table 5-5 Sequoyah Unit 2 Calculation of the ART Values for the 1/4T Location at 52 EFPY 15 Table 5-6 Sequoyah Unit 2 Calculation of the ART Values for the 3/4T Location at 52 EFPY 16 Table 5-7 Summary of the Sequoyah Unit 2 Reactor Vessel Beltline and Extended Beltline Material ART Values 17 Table 5-8 Sequoyah Unit 2 Calculation of the ART Values for the Inlet and Outlet Nozzles at 52 EFPY 18 Table 5-9 Summary of the Sequoyah Unit 2 Nozzle Material ART Values at 52 EFPY 18 Table 5-10 RTprs Calculations for Sequoyah Unit 2 Beltline and Extended Beltline Materials at 52 EFPY 19 in

PRESSURE TEMPERATURE LIMITS REPORT List of Figures Figure 2-1 Sequoyah Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 100°F/hr) Applicable for the First 44.7 EFPY (w/ Margins for Instrumentation Errors of 10°F and 60 psig) 4 Figure 2-2 Sequoyah Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First44.7 EFPY (w/ Margins for Instrumentation Errors of 10°F and 60 psig) 5 Figure 3-1 Sequoyah Unit 2 Selected LTOPS Setpoints 9 IV

PRESSURE TEMPERATURE LIMITS REPORT 1.0 RCS Pressure Temperature Limits Report (PTLR)

This PTLR for SequoyahUnit 2 has been prepared in accordance with the requirements of Technical Specification (TS) 5.6.4. Revisions to the PTLRshall be providedto the NRC after issuance.

This report affects TS 3.4.3, RCS Pressure/Temperature Limits (P/T) Limits, TS 3.4.12, Low Temperature Overpressure Protection (LTOP) System and TS 3.5.2, ECCS - Operating.

2.0 RCS Pressure and Temperature Limits The limits for TS 3.4.3 are presented in the subsections which follow and were developed using the NRC approved methodologies specified inTS 5.6.4 with exception ofASME Code Case N-640[11] (Use ofKk),

WCAP-15984-P[12] (Elimination ofthe Flange Requirement), 1996 Version ofAppendix G[4] and the revised fluences[15]. The operability requirements associated with LTOPS are specified inTS 3.4.12 and were determined to adequately protect the RCS against brittle fracture in the event of an LTOP Transient in 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°F 2.1.2 The RCS temperature rate-of-change limits are:

a. A maximum heatup rate of 100°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 10°F in any one hour period during inservice 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 criticality are specified by Figures 2-1 and 2-2. These P/T limit curves were originally documented in WCAP-15321, Revision 2[10] and were applicable to 32Effective Full-Power Years (EFPY). The applicability of the 32 EFPY P-T limit curves for Sequoyah Unit 2 was extended to 44.7 EFPY as a part of thetechnical evaluations documented in WCAP-17539-NP, Revision 1[15]. Note that the 44.7 EFPY applicability date of the Sequoyah Unit 2 P/T limit curves is less than the plant's End-of-License (EOLE) term of 52 EFPY.

2.1.4 Nozzle P/T limit curves were developed at 52 EFPY as a part of the evaluations contained in MCOE-LTR-16-12-NP[l6] to satisfy NRC Regulatory Issue Summary (RIS 2014-11)[17]. The nozzle curves are compared to the cylindrical shell P/T limit curves (developed using Reference 9 methodology) with a revised applicability of 44.7 EFPY in Reference 16, which shows that the nozzle P/T limit curves are bounded by the cylindrical shell P/T limit curves. Note that Reference 16 compares 52 EFPY nozzle P/T limit curves to 44.7 EFPY cylindrical shell P/T limit curves; this approach is conservative.

PRESSURE TEMPERATURE LIMITS REPORT 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 the subsections which follow. These lift setpoints have been developed using the NRC-approved methodologies specified in TS 5.6.4.

3.1 Pressurizer PORV Lift Setting Limits The pressurizer PORV lift setpoints are specified by Figure 3-1 and Table 3-1 (Ref. 10). The limits for the LTOPS setpoints are contained in the 44.7 EFPY steady-state curves (Table 2-2), which are beltline conditions and are not compensated for pressure differences between the pressurizer transmitter and the reactor midplane/beltline. However, these curves are adjusted for instrument inaccuracies. The pressure difference between the pressurizer transmitter and the reactor vessel midplane/beltline with four reactor coolant pumps in operation is 68.3 psi (Ref. 13).

Note: These setpoints include allowance for the pressure difference between the pressurizer transmitter and the reactor vessel midplane/beltline and the 50°F thermal transport effect for heat injection transients. Ademonstrated accuracy calculation (Reference 14) has been performed to confirm that the setpoints will maintain the system pressure within the established limits when the pressure difference between the pressure transmitter and reactor midplane and maximum temperature/pressure instrument uncertainties are applied to the setpoints.

3.2 Arming Temperature The LTOPS arming temperature is based upon the methodology defined in the Sequoyah Nuclear Plant Unit 2Technical Specifications Administrative Controls Section 5.6.4. The arming temperature shall be <

350°F.

4.0 Reactor Vessel Material Surveillance Program The reactor vessel material irradiation surveillance specimens shall be removed and examined to determine changes inmaterial properties. The removal schedule is provided inTable 4-1. The results of these examinations shall be used to update Figures 2-1,2-2 and 3-1.

The pressure vessel steel surveillance program (WCAP-8513[,]) is in compliance with Appendix Hto 10 CFR 50, "Reactor Vessel Material Surveillance Program Requirements^." The material test requirements and the acceptance standard utilize the reference nil-ductility temperature RTNDT, which isdetermined in accordance with ASTM E23[3]. The empirical relationship between RTNDT and the fracture toughness of thereactor vessel steel is developed in accordance with Code Case N-640 of Section XIof theASME Boiler and Pressure Vessel Code, Appendix G, "Fracture Toughness Criteria for Protection Against Failure[4]." The surveillance capsule removal schedule meets the requirements ofASTM E185-82[5]. The removal schedule is provided in Table 4-1.

PRESSURE TEMPERATURE LIMITS REPORT 5.0 Supplemental Data Tables Table 5-1 contains a comparison of measured surveillance material 30 ft-lb transition temperature shifts and 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 capsule data. Note that in the calculation of the surveillance weld chemistry factor, the ratio procedure from Regulatory 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 cooldown limit 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 52 EFPY for each beltline material and extended beltline material in the Sequoyah Unit 2 reactor vessel. The limiting beltline material was the Intermediate Shell Forging 05.

Table 5-7 provides a summary of the adjusted reference temperature (ART) values of the Sequoyah Unit 2 reactor vessel beltline materials and extended beltline materials at the 1/4T and 3/4T locations for 52 EFPY Table 5-8 shows the maximum ART value at 52 EFPY for each inlet and outlet nozzle in the Sequoyah Unit 2 reactor vessel.

Table 5-9 provides a summary of the limiting ART values of the Sequoyah Unit 2 inlet and outlet nozzles at 52 EFPY.

Table 5-10 provides RTPTS values for Sequoyah Unit 2 at 52 EFPY.

PRESSURE TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING05 LIMITING ART VALUES AT 44.7 EFPY: 1/4T, 142°F 3/4T, 115°F 2500 OperlimVersion:5.1 Run:5694 2250 2000 Unacceptable Acceptable Operation Operation 1750 O

CO Q. 1500 Critical Limit (ft 100 Deg. F/Hr 9 1250 2

a 3

O 1000 (Q

o 750 Critical ty Limit based on 500 inservice hydrostatic test temperature (214°F) for the Minumum service period up to 44.7 EFPY Boltup Temp 250 = 50°F 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 2-1 Sequoyah Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 100°F/hr) Applicable forthe First 44.7 EFPY(w/ Margins for Instrumentation Error of 10°F and 60 psig) (Plotted Data provided on Table 2-1)

PRESSURE TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING 05 LIMITING ART VALUES AT 44.7 EFPY: 1/4T, 142°F 3/4T, 115°F 2500 lOperlim Version:5.1 Run:5694 2250 Unacceptable Acceptable 2000 Operation Operation 1750 e>

CO 1500 3

(ft (ft

£ 1250 Q.

T5

£ J5 3 1000 O

w O

750 500 250 Minimum Boltup Temp

= 50°F 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 2-2 Sequoyah Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 44.7 EFPY (w/ Margins for Instrumentation Error of 10°F and 60 psig) (Plotted Data provided on Table 2-2)

PRESSURE TEMPERATURE LIMITS REPORT Table 2-1 Sequoyah Unit 2 Heatup Limits at 44.7 EFPY (with Uncertainties for Instrumentation Errors of 10°F and 60 psig) 100 Heatup 100 Critical Limit Leak Test Limit T P T P T P 50 0 214 0 198 2000 50 591 214 607 214 2485 55 595 214 614 60 601 214 622 65 607 214 657 70 614 214 650 75 622 214 647 80 630 214 646 85 640 214 648 90 646 214 653 95 646 214 661 100 646 214 671 105 646 214 680 110 646 214 685 115 646 214 701 120 646 214 720 125 648 214 743 130 653 214 769 135 661 214 798 140 671 215 832 145 685 220 869 150 701 225 911 155 720 230 959 160 743 235 1011 165 769 240 1069 170 798 245 1134 175 832 250 1206 180 869 255 1286 185 911 260 1374 190 959 265 1471 195 1011 270 1579 200 1069 275 1698 205 1134 280 1829 210 1206 285 1974 215 1286 290 2134 220 1374 295 2311 225 1471 230 1579 235 1698 240 1829 245 1974 250 2134 255 2311 I

PRESSURE TEMPERATURE LIMITS REPORT Table 2-2 Sequoyah Unit 2 Cooldown Limits at 44.7 EFPY (with Uncertainties for Instrumentation Errors of 10°F and 60 psig)

Steady State 20F 40F 60F 100F T P T P T P T P T P 50 0 50 0 50 0 50 0 50 0 50 591 50 552 50 503 50 461 50 366 55 595 55 554 55 508 55 466 55 372 60 601 60 558 60 514 60 470 60 380 65 607 65 564 65 521 65 478 65 389 70 614 70 572 70 529 70 486 70 399 75 622 75 580 75 538 75 496 75 410 80 630 80 589 80 548 80 506 80 423 85 640 85 599 85 559 85 518 85 437 90 650 90 610 90 571 90 531 90 453 95 661 95 623 95 584 95 546 95 470 100 674 100 636 100 599 100 562 100 490 105 688 105 652 105 616 105 580 105 512 110 703 110 668 110 634 110 600 110 536 115 720 115 687 115 654 115 622 115 563 120 739 120 707 120 676 120 647 120 593 125 760 125 730 125 701 125 674 125 626 130 783 130 755 130 729 130 704 130 663 135 809 135 783 135 759 135 738 135 704 140 837 140 814 140 793 140 775 140 749 145 868 145 848 145 831 145 816 145 800 150 902 150 885 150 872 150 862 150 856 155 940 155 927 155 918 155 913 160 982 160 973 160 968 165 1028 165 1024 170 1080 175 1136 180 1199 185 1268 190 1344 195 1429 200 1522 205 1625 210 1739 215 1865 220 2004 225 2158 230 2328

PRESSURE TEMPERATURE LIMITS REPORT Table 3-1 Selected Setpoints00, Sequoyah Unit 2 Indicated RCS PCV-456(b) PCV-455A(C)

Temperature Setpoint (psig) Setpoint (psig)

(°F) 50 497 454 73 497 454 123 567 525 173 660 616 200 715 626 327 715 626 373 715 626 500 2335 2335 Notes:

(a) From Reference 19 and confirmed per Reference 20.

(b) PCV-456 is PORV#2.

(c) PCV-455AisPORV#l.

PRESSURE TEMPERATURE LIMITS REPORT 2500

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I v-:p-^A 2000 j- ---1 -i - -

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'55 rr 1500 1

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1000 _. 4 - 1 * -[* 1 2 1 - - -- - - l -- .-. -j -- (

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""" J l i 500 PCV456 _Hl 1 .

i PCV455A -1 1 i j -- -l - - - ;-- -- - - j i 1 1 1  ! 1 j 1 1 50 100 150 200 250 300 350 400 450 500 550 RCSTemperature (°F)

Figure 3-1 Sequoyah Unit 2 - COMS PORV Setpoints vs. Indicated RCS TemperatureI191 (Includes Pressure and Temperature Instrument Uncertainties. Plotted Data provided on Table 3-1)

PRESSURE TEMPERATURE LIMITS REPORT Table 4-1 Sequoyah Unit 2 Reactor Vessel Surveillance Capsule Withdrawal Schedule Removal Time Fluence(0 Capsule Location Lead Factor'" (EFPY)(b) (n/cm2,E>1.0 MeV)(a)

T 40° 3.11 1.07 2.44xlOl8(c)

U 140° 3.17 2.91 6.54xlOl8(c)

X 220° 3.18 5.36 1.16xl0,9(c)

Y 320° 3.15 10.55 2.02xl0,9(c,d)

W 140°(184°) 3.18 >33.7 2.57xl019(e)

S 40° (4°) 3.11 Note (f) Note (f)

V 176° 0.94 Standby (g) z 356° 0.94 Standby (g)

Notes:

(a) Updated in the time-limiting aging analysis (TLAA) fluence evaluation (WCAP-17539-NP Revision 1[15]).

(b) Effective Full Power Years (EFPY) from plant startup.

(c) Plant specific evaluation.

(d) This fluence is not less than once orgreater than twice the peak end oflicense (32 EFPY) fluence.

(e) Capsule Wwas relocated from the 184° location to the 140° location. This capsule should beremoved with afluence value corresponding to at least one times the 60-year EOL vessel fluence (2.57 x 1019n/cm2) but less than two times the 60-year EOL vessel fluence (5.14 x 1019n/cm2). Capsule Wwill reach a fluence of 2.57 x 1019 (E > 1.0 MeV) at 33.7 EFPY, at the 140° position.[18]

(f) Capsule S was relocated from the 4° location to the 40° location. Capsule S was discovered to have lifted 2 inches from its 40° azimuthal basket atEOC 20. This small elevation change inside the basket had no impact on fluence exposure. The capsule was removed at EOC 20 and stored in the spent fuel pool. The capsule can be reinserted in an alternate, higher lead factor location in the vessel if additional material data is needed in support ofa second plant license renewal for 80 years oflicensed operation.[18]

(g) Capsules Vand Z lag the vessel fluence and are not predicted toreach 60-year EOL fluence before EOLE.

Capsules Vand Z should remain in the reactor and continue toaccrue irradiation for potential future testing, if needed.

10

PRESSURE TEMPERATURE LIMITS REPORT Table 5-1 Comparison of the Sequoyah Unit 2 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions 30 ft-lb Transition Upper Shelf Energy Temperature Shift Decrease Material Capsule Fluence Predicted Measured Predicted Measured (x 1019 n/cm2) (°F)(a) (°F)(b) (%)(a) (%)(c)

Intermediate Shell T 0.244 60.33 63.65 17 12 Forging 05 0.654 85.22 79.31 21 16 U

(Tangential)

X 1.16 100.23 85.7 23 8 (Heat # 288757 /

981057) Y 2.02 114.67 134.12 26 22 Intermediate Shell T 0.244 60.33 48.73 17 7 Forging 05 U 0.654 85.22 66.06 21 9 (Axial)

X 1.16 100.23 110.04 23 2 (Heat # 288757 /

981057) Y 2.02 114.67 89.21 26 22 Weld Metal T 0.244 43.12 74.56 20 2 (Heat # 4278)(d) U 0.654 60.91 130.38 25 6 X 1.16 71.63 44.22 29 35 Y 2.02 81.96 86.91 33 3 HAZ Metal T 0.244 - - 24.58 - - 2 U 0.654 - - 64.03 - - 14 X 1.16 - - 28.29 - - 19 Y 2.02 - -

50.32 - - 39 Notes:

(a) Based on Regulatory Guide 1.99, Revision 2, methodology using the mean weight percent values of copper 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 El 85-82.

(d) Surveillance Weld was fabricated from weld wire type SMIT 89, Heat # 4278, Flux type SMIT 89, Lot # 1211.

11

PRESSURE TEMPERATURE LIMITS REPORT Table 5-2 Calculation ofChemistry Factors using Sequoyah Unit 2 Surveillance Capsule Data Material Capsule Capsule f°° FF(b) ARTNDT(C) FF*ARTNDT FF2 Intermediate Shell T 2.44E+18 0.618 63.65 39.33 0.382 Forging 05 U 6.54E+18 0.881 79.31 69.87 0.776 (Tangential)

X 1.16E+19 1.041 85.70 89.25 (Heat #288757/ 1.085 981057) Y 2.02E+19 1.192 134.12 159.83 1.420 Intermediate Shell T 2.44E+18 0.618 48.73 30.11 0.382 Forging 05 U 6.54E+18 0.881 66.06 58.20 0.776 (Axial)

X 1.16E+19 1.041 110.04 114.60 (Heat # 288757 / 1.085 981057) Y 2.02E+19 1.192 89.21 106.31 1.420 SUM: 667.51°F 7.325 CF05 =Z(FF*RTNDT) + S(FF2) =(667.51)+ (7.325) =91.1°F Surveillance Weld T 2.44E+18 0.618 69.34 (74.56) 42.85 0.382 Material^

U 6.54E+18 0.881 121.25(130.38) 106.82 0.776 (Heat#4278)(e)

X 1.16E+19 1.041 41.12(44.22) 42.83 1.085 Y 2.02E+19 1.192 80.83 (86.91) 96.32 1.420 SUM: 288.82°F 3.663 CF surv. weid =I(FF

  • RTNDT) + S( FF2) =(288.82°F) + (3.663) =78.9°F (a) f=Calculated fluence (n/cm2, E>1.0 MeV), updated per TLAA, WCAP-17539-NP, Revision l[l5]

(b) FF= fluence factor= f(028-°A*l°^ .

(c) ARTNDT values are the measured 30 ft-lb shift values taken from Table 5-10 of WCAP-15320[7].

(d) The surveillance weld metal ARTNDT values have been adjusted by aratio factor of0.93.

(e) Surveillance Weld was fabricated from weld wire type SMIT 89, Heat #4278, Flux type SMIT 89 Lot #

1211.

12

PRESSURE TEMPERATURE LIMITS REPORT Table 5-3 Reactor Vessel Beltline and Extended Beltline Material Unirradiated Toughness Properties for Sequoyah Unit 2 Fracture Toughness Chemical Composition Properties Initial Material Description Initial Upper-Cu Ni P Mn dt (a) Shelf KIndt (Wt. %) (Wt. %) (Wt. %) (Wt. %) Energy

(°F)

(ft-lb)

Reactor Vessel Beltline Materials(c)

Intermediate Shell Forging 05 0.13 0.76 0.014 0.70 10 93 (Heat #288757/981057)

Lower Shell Forging 04 0.14 0.76 0.012 0.68 -22 100 (Heat # 990469 / 293323)

Intermediate to Lower Shell Forging Circumferential Weld Seam W05(b) 0.12 0.11 0.016 1.50 -4 102 (Heat # 4278)

Surveillance Weld(b) 0.13 0.11 0.016 1.50 Reactor Vessel Extended Beltline Materials(c)

Upper Shell Forging 06 0.16 0.84 0.016 0.72 5 68 (Heat #981201/285849)

Bottom Head Ring 03 0.16 0.77 0.016 0.73 5 64 (Heat #981177/288872)

Upper Shell to Intermediate Shell 0.08 1.0 0.019 1.52 10 78 Circumferential Weld W06 (Heat #721858)

Lower Shell to Bottom Head Ring 0.08 1.0 0.019 1.52 10 78 Weld W04 (Heat #721858)

Notes:

(a) The Initial RTNDT values are measured values.

(b) Circumferential Weld Seam was fabricated with weld wire type SMIT 89, Heat # 4278, Flux type SMIT 89, Lot # 1211 and is representative ofthe intermediate to lower shell circumferential weld.

(c) Chemical Composition and Fracture Toughness Properties were updated and Extended Beltline materials were added per WCAP-17539-NP, Revision 1, Table 3-2[,5].

13

PRESSURE TEMPERATURE LIMITS REPORT Table 5-4 Peak Neutron Fluence Projections on the Reactor Vessel Clad/Base Metal Interface for Sequoyah Unit 2 (xl019n/cm2, E> 1.0 MeV)

Fluence(a)

Reactor Vessel Material 32 EFPY 52 EFPY Upper Shell Forging06 0.0342 0.0552 Intermediate Shell Forging 05 1.66 2.57 Lower Shell Forging04 1.66 2.57 Bottom Head Ring 03 0.200 0.316 Upper Shell to Intermediate Shell 0.0342 0.0552 Circumferential Weld W06 Intermediate Shell to Lower Shell 1.65 2.55 Circumferential Weld W05 Lower Shell to Bottom Head Ring Weld 0.200 0.316 W04 Note:

(a) Fluence was updated per TLAA, WCAP-17539-NP, Revision 1 Tables 2-3 and 2-4[15).

14

PRESSURE TEMPERATURE LIMITS REPORT Table 5-5 Sequoyah Unit 2 Calculation of the ARTValues for the 1/4T Location at 52 EFPY(a)

IRTndt ARTNDT RG 1.99 CF ) ) Margin(d) ART(e)

Material FF R2 Method (°F) (°F) (°F)

(°F) (°F)

Reactor Vessel Beltline Materials Position 95.0 1.1208 10 106.5 34.0 150.5 1.1 Intermediate Shell Forging 05 Position 91.1 1.1208 10 102.1 17.0(f) 129.1 2.1 Position Lower Shell Forging 04 104.0 1.1208 -22 116.6 34.0 128.6 1.1 Position 63.0 1.1187 -4 70.5 56.0 122.5 Intermediate to Lower Shell 1.1 Circumferential Weld Seam Position 78.9 1.1187 -4 88.3 56.0(f) 140.3 2.1 Reactor Vessel Extended Beltline Materials Position Upper Shell Forging 06 123.4 0.2331 5 28.8 28.8 62.5 1.1 Position Bottom Head Ring 03 122.3 0.5576 5 68.2 34.0 107.2 1.1 Upper Shell to Intermediate Shell Circumferential Weld Position 108.0 0.2331 10 25.2 25.2 60.4 W06 1.1 (Heat #721858)

Lower Shell to Bottom Head Position Ring Weld W04 (Heat 108.0 0.5576 10 60.2 56.0(f) 126.2 1.1

  1. 721858)

Notes:

(a) Neutron fluence values used for all materials obtained from Table 5-4 for 52 EFPY.

(b) Initial RTNDT values are measured values.

(c) ARTNDT = CF*FF (d) Margin =2 *(gl2 + aA2),/2 (e) ART = Initial RTNDT + ARTNDT + Margin (°F)

(f) Per Reference 15, the surveillance forging data is deemed credible, and the surveillance weld material is deemed non-credible. Thus, the full aA will be used to determine margin for the surveillance weld, and the reduced aA will be used to determine margin for the surveillance forging.

15

PRESSURE TEMPERATURE LIMITS REPORT Table 5-6 Sequoyah Unit 2 Calculation of the ART Values for the 3/4T Location at 52 EFPY(a)

Material RG 1.99 CF FF IRTndt ARTNDT( Margin(d ART(e)

R2 ) c)

(°F) )

(°F)

Method (°F) (°F) (°F)

Reactor Vessel Beltline Materials Position 95.0 0.8386 10 79.7 34.0 123.7 1.1 Intermediate Shell Forging 05 Position 91.1 0.8386 10 76.4 17.0(f) 103.4 2.1 Position Lower Shell Forging 04 104.0 0.8386 -22 87.2 34.0 99.2 1.1 Position 63.0 0.8364 -4 52.7 52.7 101.4 Intermediate to Lower Shell 1.1 Circumferential Weld Seam Position 2.1 78.9 0.8364 -4 66.0 56.0(f) 118.0 Reactor Vessel Extended Beltline Materials Position Upper Shell Forging 06 123.4 0.1244 5 15.3 15.3 35.7 1.1 Position Bottom Head Ring 03 122.3 0.3469 5 42.4 34.0 81.4 1.1 Upper Shell to Intermediate Shell Circumferential Weld Position 108.0 0.1244 10 13.4 13.4 36.9 W06 1.1 (Heat #721858)

Lower Shell to Bottom Head Position Ring Weld W04 (Heat 108.0 0.3469 10 37.5 37.5 84.9 1.1

  1. 721858) fotes: '

(a) Neutron fluence valuesused for all materials obtained from Table5-4 for 52 EFPY.

(b) Initial RTNDT values are measured values.

(c) ARTNDT = CF*FF (d) Margin = 2 *(g;2 + aA2)1/2 (e) ART= Initial RTNDT + ARTNDT + Margin (°F)

(f) Per Reference 15, the surveillance forging data isdeemed credible, and the surveillance weld material is deemed non-credible. Thus, the full aA will be used to determine margin for the surveillance weld, and the reduced aA will be used to determine margin for the surveillance forging.

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PRESSURE TEMPERATURE LIMITS REPORT Table 5-7 Summary of the Sequoyah Unit 2 Reactor Vessel Beltline and Extended Beltline Material ART Values Material RG1.99R2 1/4 ART 3/4 ART Method (°F) (°F)

Reactor Vessel Beltline Materials Position 1.1 150.5 123.7 Intermediate Shell Forging 05 Position 2.1 129.1 103.4 Lower Shell Forging 04 Position 1.1 128.6 99.2 Intermediate to Lower Shell Position 1.1 122.5 101.4 Circumferential Weld Seam Position 2.1 140.3 118.0 Reactor Vessel Extended Beltline Materials Upper Shell Forging 06 Position 1.1 62.5 35.7 Bottom Head Ring 03 Position 1.1 107.2 81.4 Upper Shell to Intermediate 60.4 36.9 Position 1.1 Shell Circumferential Weld Lower Shell to Bottom Head Position 1.1 126.2 84.9 Ring Weld 17

PRESSURE TEMPERATURE LIMITS REPORT Table 5-8 Sequoyah Unit 2 Calculation ofthe ART Values for the Inlet and Outlet Nozzles at 52 EFPY(a)

Fluence Material Heat# (xlO,9n/cm2, IRTNDT ARTNDT Margin(c) ART(d)

(°F) (°F) (°F) (°F)

E> 1.0 MeV)

Inlet Nozzle 11 4868 0.00340 1 0.00 0.0 1 Inlet Nozzle 12 4872 0.00340 -22 0.00 0.0 -22 Inlet Nozzle 13 4877 0.00340 -31 0.00 0.0 -31 Inlet Nozzle 14(e) 4886 0.00340 -31 0.00 0.0 -31 Outlet Nozzle 15 4867 0.00340 -21 0.00 0.0 -21 Outlet Nozzle 16 4873 0.00340 -19 0.00 0.0 -19 Outlet Nozzle 17 4878 0.00340 -40 0.00 0.0 -40 Outlet Nozzle 18 4887 0.00340 -22 0.00 0.0 -22 Notes:

(a) Neutron fluence value for the inlet nozzle to upper shell welds at 52 EFPY was used for all nozzle materials per MCOE-LTR-16-12-NP[16].

(b) Initial RTNDT values are measured values.

(c) Margin =2*(g;2 + aA2)1/2 =0for materials with measured IRTNDT values and fluence values less than 1x 1017 n/cm2 (E >1.0 MeV).

(d) ART =Initial RTNDT +ART^t +Margin (°F); ARTNDT =0 for fluence values less than 1x 1017 n/cm2 (E >1.0 MeV) per Reference 18.

(e) TNDT =IRTNDT for this nozzle. This value was updated from that in the Sequoyah Unit 2 Technical Specification Bases per MCOE-LTR-16-12-NP[16].

Table 5-9 Summary of the Sequoyah Unit 2 Nozzle Material ARTValues at 52 EFPY Nozzle Material LimitingART Value (°F)

Inlet Nozzle 11 Heat # 4868 1 Outlet Nozzle 16 Heat # 4873 -19 18

PRESSURE TEMPERATURE LIMITS REPORT Table 5-10 RTp-rs Calculations for Sequoyah Unit 2 Beltline and Extended Beltline Materials at 52 EFPY(a)

Material Fluence FF CF ARTpts Margin PT K-*NDT(U)

(°) PT W (x 10 ,9 (°F) (°F) (°F) (°F) (°F) n/cm2, E>1.0 MeV)

Reactor Vessel Beltline Materials Intermediate Shell Forging 05 2.57 1.2531 95.0 119.0 34.0 10 163.0 Intermediate Shell Forging 05 2.57 1.2531 91.1 114.2 17.0(e) 10 141.2 (Using S/C Data)

Lower Shell Forging 04 2.57 1.2531 104.0 130.3 34.0 -22 142.3 Circumferential Weld Metal 2.55 1.2511 63.0 78.8 56.0 -4 130.8 Circumferential Weld Metal 2.55 1.2511 78.9 98.7 56.0(e) -4 150.7 (Using S/C Data)

Reactor Vessel Extended Beltline Materials Upper Shell Forging 06 0.0552 0.3087 123.4 38.1 34.0 5 77.1 Bottom Head Ring 03 0.316 0.6837 122.3 83.6 34.0 5 122.6 Upper Shell to Intermediate Shell 0.0552 0.3087 108.0 33.3 33.3 10 76.7 Circumferential Weld W06 Lower Shell to Bottom Head 0.316 0.6837 108.0 73.8 56.0(e) 10 139.8 Ring Weld W04 Notes:

(a) Neutron fluence values used for all materials obtained from Table 5-4 for 52 EFPY.

(b) ARTpts = CF

  • FF (c) Initial RTndt values are measured values (d) RTpts = RTndt(U) + ARTpts + Margin (°F)

(e) Per Reference 15, the surveillance forging data is deemed credible, and the surveillance weld material is deemed non-credible. Thus, the full aA will be used to determine margin for the surveillance weld, and the reduced aA will be used to determine margin for the surveillance forging.

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PRESSURE TEMPERATURE LIMITS REPORT 6.0 References

1. WCAP-8513, Tennessee Valley Authority Sequoyah Unit No. 2 Reactor Vessel Radiation Surveillance Program, J. A. Davidson, et. al., November 1975.
2. Code of Federal Regulations, 10CFR50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C.
3. ASTM E23 Standard Test Method Notched Bar Impact Testing ofMetallic Materials, in ASTM Standards, American Society for Testing and Materials, Philadelphia, PA.
4. Section XI ofthe ASME Boiler and Pressure Vessel Code, Appendix G, Fracture Toughness Criteriafor Protection Against Failure
5. ASTM El85-82, Annual Book ofASTM Standards, Section 12, Volume 12.02, Standard Practice for Conducting Surveillance Testsfor Light-Water Cooled Nuclear Power Reactor Vessels.
6. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement ofReactor Vessel Materials, U.S.

Nuclear Regulatory Commission, May 1988.

7a. WCAP-15320, Analysis ofCapsule Yfrom the Tennessee Valley Authority Sequoyah Unit 2 Reactor Vessel Radiation Surveillance Program, T.J. Laubham, et. al., November 1999.

7b. WCAP-10509, Analysis ofCapsule Tfrom the Tennessee Valley Authority Sequoyah Unit 2 Reactor Vessel Radiation Surveillance Program, R. S. Boggs, etal, April 1984.

7c. Southwest Research Institute Nondestructive Evaluation Science and Technology Division, Reactor Vessel Material Surveillance Program and Technology Division, Reactor Vessel Material Surveillance Programfor Sequoyah Unit 2: Analysis ofCapsule U, Final Report SwRI Project 17-8851 TVA Contract 85PJH-964430, January 1990.

7d. WCAP-13545, Analysis ofCapsule Xfrom the Tennessee Valley Authority Sequoyah Unit 2 Reactor 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 ATI Consulting, March 1999.
9. WCAP-14040-NP-A, Revision 2, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, J.D. Andrachek, et. al January 1996.
10. WCAP-15321, Revision 2, Sequoyah Unit 2Heatup and Cooldown Limit Curvesfor Normal Operation and PTLR Support Documentation, J.H. Ledger, et.al., July 2003.
11. ASME Code Case N-640, Alternative Reference Fracture Toughnessfor Development ofP-T Limit Curvesfor Section XI, Division 1, dated February 26, 1999.

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PRESSURE TEMPERATURE LIMITS REPORT

12. WCAP-15984-P, Revision 01, Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation For Sequoyah Units 1 and 2, W. Bamford, etal., April 2003.
13. Westinghouse Letter to TVA, TVA-93-105, Cold Overpressure Mitigation System Code Case and Delta-P Calculation, dated May 19, 1993.
14. Calculation SQN-IC-014, DemonstratedAccuracy Calculationfor Cold Overpressure Protection System.
15. WCAP-17539-NP, Revision 1, Sequoyah Units 1 and 2 Time-LimitedAging Analysis on Reactor VesselIntegrity, E.J. Long, May 2015.
16. MCOE-LTR-16-12-NP, Revision 0, Pressure-Temperature Limit Curve Developmentfor ExtendedPlant Operationfor Sequoyah Units 1 and 2 Reactor Vessel Nozzles, E.M. Ruminski, A.M. Carolan, August 2016.
17. U.S. NRC Regulatory Issue Summary (RIS) 2014-11, Information on Licensing Applicationsfor Fracture Toughness Requirementsfor Ferritic Reactor Coolant Pressure Boundary Components, dated October 14, 2014.
18. MCOE-LTR-15-121-NP, Revision 0, Justificationfor Removal ofSurveillance Capsule Sfrom the Sequoyah Unit2 Reactor Vessel at EOC 20 and Reinsertion AfterEOC 21 with Regards to Subsequent License Renewal, E.J. Long, December 2015.
19. LTR-SCS-16-25, Revision 0, Cold Overpressure Mitigation System (COMS) Setpoint Analysis for Sequoyah Units 1 and 2 PORVReplacement, R.A. Freund, B.D. Jaskiewicz, November 2016.
20. LTR-SCS-17-43, Revision 0, Sequoyah LTOPSSetpoint Implementationfor PORV Replacement Program Delay, R.A. Freund, September 2017.

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