ML18136A495

Pressure Temperature Limits Report, Revision 6

ML18136A495
Person / Time
Site:	Sequoyah
Issue date:	05/15/2018
From:	Anthony Williams Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML18136A495 (32)

Text

Tennessee Valley Authority, Sequoyah Nuclear Plant, P.O. Box 2000, Soddy Daisy, Tennessee 37384 May 15, 2018 10CFR50.4 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Sequoyah Nuclear Plant, Unit 1 Renewed Facility Operating License Nos. DPR-77 NRC Docket Nos. 50-327

Subject:

Sequoyah Unit 1 Pressure Temperature Limits Report, Revision 6 In accordance with Sequoyah Nuclear Plant (SQN) Unit 1 Technical Specification (TS) 5.6.4.C, enclosed is the Pressure Temperature Limits Report (PTLR), Revision 6. The 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 years (EFPY) to 52 EFPY; to document the evaluation of the reactor vessel materials in the extended beltline region; and to provide a 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 April 27, 2018, during the Unit 1 Cycle 22 Refueling Outage. Associated administrative changes were made for report consistency.

There are no new regulatory commitments in this letter. If you have any questions, please contact Michael McBrearty, SQN Site Licensing Manager at (423) 843-7170.

Anthony L. Williams Site Vice President Sequoyah Nuclear Plant Enclosure Sequoyah Unit 1 Pressure Temperature Limits Report, Revision 6 printed on recycled paper

U.S. Nuclear Regulatory Commission Page 2 May 15, 2018 ZTK:STB Enclosure cc (Enclosure):

NRC Regional Administrator - Region II NRC Senior Resident Inspector - SQN

ENCLOSURE SEQUOYAH UNIT 1 PRESSURE TEMPERATURE LIMITS REPORT, REVISION 6

PRESSURE TEMPERATURE LIMITS REPORT

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Letter No. Nl0810 Dale: April 30, 2018 *...;.._.

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SQEP (~!) BY: W. Chris Reneau Tennessee Valley Authority Sequoyah Unit 1 Pressure Temperature Limits Report Revision 6, September 2017 PROJECT Sequoyah DISCIPLINE N CONTRACT 4411 UNIT_ _.:.1_ _

DESC. RCS Pressure-Temperature Limit Report DWG/DOC NO._..:...P=TL=R=-.;_1_ _ _ _ _ _ __

SHEET OF REV. 06 DATE 04/30/18 ECN/DCN FILE N2N-081 EDMS, WT CA-K

PRESSURE TEMPERATURE LIMITS REPORT

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Contractor li-0111 any port ol his r -

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Letter No. Nl0810 Dale: April 30, 2018 *...;.._.

~F.NS§E -.a.ei~ -

SQEP (~!) BY: W. Chris Reneau Tennessee Valley Authority Sequoyah Unit 1 Pressure Temperature Limits Report Revision 6, September 2017 PROJECT Sequoyah DISCIPLINE N CONTRACT 4411 UNIT_ _.:.1_ _

DESC. RCS Pressure-Temperature Limit Report DWG/DOC NO._..:...P=TL=R=-.;_1_ _ _ _ _ _ __

SHEET OF REV. 06 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 ............................................................................................................................ 1 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 ..................................................................................................................................... 23 ii

PRESSURE TEMPERATURE LIMITS REPORT List of Tables Table 2-1 Sequoyah Unit 1 Heatup Limits at 52 EFPY (with Uncertainties for Instrumentation Errors of 10°F and 60 psig) ........................................... 6 Table 2-2 Sequoyah Unit 1 Cooldown Limits at 52 EFPY (with Uncertainties for Instrumentation Errors of 10°F and 60 psig) ........................................... 8 Table 3-1 Selected Setpoints, Sequoyah Unit 1 ................................................................................ 10 Table 4-1 Sequoyah Unit 1 Reactor Vessel Surveillance Capsule Withdrawal Schedule ................. 12 Table 5-1 Comparison of the Sequoyah Unit 1 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, revision 2, Predictions....................................................................................................... 13 Table 5-2 Calculation of Chemistry Factors using Sequoyah Unit 1 Surveillance Capsule Data..... 14 Table 5-3 Reactor Vessel Beltline and Extended Beltline Material Unirradiated Toughness Properties for Sequoyah Unit 1 ......................................................................................... 15 Table 5-4 Peak Neutron Fluence Projections on the Reactor Vessel Clad/Base Metal Interface for Sequoyah Unit 1 (x 1019 n/cm2, E > 1.0 MeV) ................................................................. 17 Table 5-5 Sequoyah Unit 1 Calculation of the ART Values for the 1/4T Location at 52 EFPY ....... 18 Table 5-6 Sequoyah Unit 1 Calculation of the ART Values for the 3/4T Location at 52 EFPY ....... 19 Table 5-7 Summary of the Sequoyah Unit 1 Reactor Vessel Beltline and Extended Beltline Material ART Values ....................................................................................................................... 20 Table 5-8 Sequoyah Unit 1 Calculation of the ART Values for the Inlet and Outlet Nozzles at 52 EFPY.. ........................................................................................................... 21 Table 5-9 Summary of the Sequoyah Unit 1 Nozzle Material ART Values at 52 EFPY... 21 Table 5-10 RTPTS Calculations for Sequoyah Unit 1 Beltline and Extended Beltline Materials at 52 EFPY...... .......................................................................................................................... 22 iii

PRESSURE TEMPERATURE LIMITS REPORT List of Figures Figure 2-1 Sequoyah Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 100°F/hr) Applicable for the First 52 EFPY (w/ Margins for Instrumentation Errors of 10°F and 60 psig) ............................................ 4 Figure 2-2 Sequoyah Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 52 EFPY (w/ Margins for Instrumentation Errors of 10°F and 60 psig) ............................................ 5 Figure 3-1 Sequoyah Unit 1 Selected LTOPS Setpoints .................................................................... 11 iv

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

This PTLR for Sequoyah Unit 1 has been prepared in accordance with the requirements of Technical Specification (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, TS 3.4.12, Low Temperature Over Pressure 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 in TS 5.6.4 with exception of ASME Code Case N-640[13] (Use of KIc),

WCAP-15984-P[14] (Elimination of the Flange Requirement), 1996 Version of Appendix G[4] and the revised fluences[15]. The operability requirements associated with LTOPS are specified in TS 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-15293, Revision 2[10] and were applicable to 32 Effective Full-Power Years (EFPY). The applicability of the 32 EFPY P-T limit curves for Sequoyah Unit 1 was extended to 52 EFPY as a part of the technical evaluations documented in WCAP-17539-NP, Revision 1[15].

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[16] 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 52 EFPY in Reference 16, which shows that the nozzle P/T limit curves are bounded by the cylindrical shell P/T limit curves.

1

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 52 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. 11).

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. A demonstrated accuracy calculation (Reference 12) 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 1 Technical 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 in material properties. The removal schedule is provided in Table 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-8233[1]) is in compliance with Appendix H to 10 CFR 50, Reactor Vessel Material Surveillance Program Requirements[2]. The material test requirements and the acceptance standard utilize the reference nil-ductility temperature RTNDT, which is determined in accordance with ASTM E23[3]. The empirical relationship between RTNDT and the fracture toughness of the reactor vessel steel is developed in accordance with Code Case N-640 of Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G, Fracture Toughness Criteria for Protection Against Failure[4]. The surveillance capsule removal schedule meets the requirements of ASTM E185-82[5]. The removal schedule is provided in Table 4-1.

2

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 0.90.

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 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 1 reactor vessel. The limiting beltline material was the Lower Shell Forging 04.

Table 5-7 provides a summary of the adjusted reference temperature (ART) values of the Sequoyah Unit 1 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 1 reactor vessel.

Table 5-9 provides a summary of the limiting ART values of the Sequoyah Unit 1 inlet and outlet nozzles at 52 EFPY.

Table 5-10 provides RTPTS values for Sequoyah Unit 1 at 52 EFPY.

3

PRESSURE TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART VALUES AT 52 EFPY: 1/4T, 216°F 3/4T, 186°F 2500 Operlim Version:5.1 Run:15680 2250 Leak Test Limit Unacceptable Acceptable 2000 Operation Operation 1750 Heatup Rate Critical Limit Calculated Pressure (PSIG) 100 Deg. F/Hr 100 Deg. F/Hr 1500 1250 1000 750 Criticality Limit based on inservice hydrostatic test 500 temperature (288°F) for the service period up to 52 EFPY Minimum 250 Boltup Temp = 50°F 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 2-1 Sequoyah Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 100°F/hr)

Applicable for the First 52 EFPY (w/Margins for Instrumentation Error of 10°F and 60 psig)

(Plotted Data provided on Table 2-1) 4

PRESSURE TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART VALUES AT 52 EFPY: 1/4T, 216°F 3/4T, 186°F 2500 Operlim Version:5.1 Run:15680 2250 Acceptable 2000 Unacceptable Operation Operation 1750 Calculated Pressure (PSIG) 1500 1250 Cooldown Rates 1000 F/Hr steady-state

-20

-40 750 -60

-100 500 250 Minimum Boltup Temp =

50°F 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 2-2 Sequoyah Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 52 EFPY (w/Margins for Instrumentation Error of 10°F and 60 psig) (Plotted Data provided on Table 2-2) 5

PRESSURE TEMPERATURE LIMITS REPORT Table 2-1 Sequoyah Unit 1 Heatup Limits at 52 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 0 288 0 272 2000 50 477 288 477 288 2485 55 477 288 477 60 477 288 477 65 477 288 477 70 477 288 478 75 477 288 478 80 477 288 480 85 477 288 481 90 477 288 483 95 477 288 485 100 477 288 487 105 477 288 490 110 477 288 493 115 477 288 497 120 477 288 500 125 477 288 505 130 477 288 508 135 477 288 515 140 477 288 517 145 477 288 527 150 477 288 528 155 478 288 541 160 480 288 541 165 483 288 555 170 487 288 557 175 493 288 571 180 500 288 575 185 508 288 589 190 517 288 609 195 528 288 631 200 541 288 656 205 555 288 684 210 571 288 714 215 589 288 748 220 609 290 786 225 631 295 828 6

PRESSURE TEMPERATURE LIMITS REPORT Table 2 (Continued)

Sequoyah Unit 1 Heatup Limits at 52 EFPY (with Uncertainties for Instrumentation Errors of 10°F and 60 psig) 100 Heatup 100 Critical Limit T P T P 230 656 300 874 235 684 305 925 240 714 310 981 245 748 315 1044 250 786 320 1112 255 828 325 1188 260 874 330 1272 265 925 335 1364 270 981 340 1466 275 1044 345 1578 280 1112 350 1702 285 1188 355 1838 290 1272 360 1988 295 1364 365 2154 300 1466 370 2337 305 1578 310 1702 315 1838 320 1988 325 2154 330 2337 7

PRESSURE TEMPERATURE LIMITS REPORT Table 2-2 Sequoyah Unit 1 Cooldown Limits at 52 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 50 552 50 503 50 457 50 408 50 305 55 553 55 505 55 458 55 409 55 306 60 555 60 507 60 459 60 410 60 307 65 556 65 509 65 460 65 411 65 308 70 558 70 510 70 462 70 412 70 309 75 560 75 512 75 464 75 414 75 311 80 561 80 514 80 465 80 416 80 313 85 564 85 516 85 468 85 418 85 315 90 566 90 518 90 470 90 420 90 318 95 569 95 521 95 473 95 423 95 321 100 571 100 524 100 476 100 426 100 325 105 575 105 527 105 479 105 430 105 329 110 578 110 531 110 483 110 434 110 333 115 582 115 535 115 487 115 438 115 338 120 586 120 540 120 492 120 443 120 344 125 591 125 545 125 497 125 449 125 351 130 596 130 550 130 503 130 456 130 358 135 602 135 556 135 510 135 463 135 367 140 608 140 563 140 517 140 471 140 376 145 616 145 571 145 525 145 479 145 387 150 623 150 579 150 534 150 489 150 399 155 632 155 588 155 544 155 500 155 412 160 642 160 599 160 556 160 512 160 427 165 652 165 610 165 568 165 526 165 443 170 664 170 623 170 582 170 541 170 461 175 677 175 637 175 597 175 558 175 482 180 691 180 652 180 614 180 577 180 505 185 707 185 669 185 633 185 597 185 530 190 724 190 688 190 654 190 620 190 558 195 743 195 709 195 677 195 646 195 590 200 764 200 733 200 702 200 674 200 624 205 788 205 759 205 731 205 705 205 663 210 814 210 787 210 762 210 740 210 706 215 843 215 819 215 797 215 779 215 754 8

PRESSURE TEMPERATURE LIMITS REPORT Table 2 (Continued)

Sequoyah Unit 1 Cooldown Limits at 52 EFPY (with Uncertainties for Instrumentation Errors)

Steady State 20F 40F 60F 100F T P T P T P T P T P 220 874 220 853 220 836 220 821 220 806 225 909 225 892 225 878 225 869 225 865 230 948 230 935 230 925 230 921 235 991 235 982 235 978 240 1038 240 1034 245 1090 250 1148 255 1212 260 1283 265 1360 270 1447 275 1542 280 1647 285 1763 290 1892 295 2034 300 2191 305 2364 9

PRESSURE TEMPERATURE LIMITS REPORT Table 3-1 Selected Setpoints(a),Sequoyah Unit 1 Indicated RCS PCV-456(b) PCV-455A(c)

Temperature Setpoint (psig) Setpoint (psig)

(ºF) 50 450 406 73 450 406 123 474 430 148 498 455 223 584 525 277 715 626 373 715 626 500 2335 2335 Notes:

(a) From Reference 18 and confirmed per Reference 19.

(b) PCV-456 is PORV#2.

(c) PCV-455A is PORV#1.

10

PRESSURE TEMPERATURE LIMITS REPORT Figure 3-1 Sequoyah Unit 1 - COMS PORV Setpoints vs. Indicated RCS Temperature [18]

(Includes Pressure and Temperature Instrument Uncertainties. Plotted Data provided on Table 3-1) 11

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

T 40° 3.15 1.07 2.41 x 1018 (c)

U 140° 3.23 2.85 6.93 x 1018 (c)

X 220° 3.22 5.26 1.16 x 1019 (c)

Y 320° 3.18 10.02 1.97 x 1019 (c,d)

S 4° 0.90 (e) (e)

W 184° 0.90 (e) (e)

V 176° 0.90 (f) (f)

Z 356° 0.90 (g) (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 or greater than twice the peak end of license (32 EFPY) fluence.

(e) Capsules S and W, which were relocated during the 1R19 outage, were found to be severely damaged during the 1R20 outage. Thus, they are no longer available to be used to satisfy the requirements of 10 CFR 50, Appendix H with consideration of license renewal to 60 total years of plant operation.

(f) Capsules V and Z are the only capsules currently remaining in the Sequoyah Unit 1 reactor vessel. Either Capsule V or Z should be withdrawn so that the capsule fluence corresponds to at least one times the 60-year EOL vessel fluence (2.66 x 1019 n/cm2), but less than two times the 60-year EOL vessel fluence (5.32 x 1019 n/cm2). However, neither of these two remaining capsules are predicted to experience a neutron fluence of 2.66 x 1019 n/cm2 prior to EOLE in their current locations. Therefore, TVA has elected to relocate Capsule V (or Capsule Z, which is radiologically equivalent to Capsule V, and could be alternatively selected) to a higher lead factor location, specifically the former Capsule U Location (140°), in order to achieve higher capsule fluence data. Note that Capsule V or Capsule Z could alternatively be moved to any open 40° capsule location, as they are all radiologically equivalent. Assuming Capsule V was relocated at the end of cycle 21, 22, or 23, the EFPY that corresponds to the time when the capsule experiences the peak EOLE vessel fluence value (2.66 x 1019 n/cm2) is approximately 35.4, 36.4, or 37.3 EFPY, respectively. See Appendix B of Reference 15 for further details on capsule relocation recommendations[15].

(g) TVA has elected, for now, to leave Capsule Z in the reactor vessel at its current 356° capsule location. This capsule could be relocated to a higher lead factor location in the future if additional metallurgical data is needed in support of a potential second license renewal to 80 total years of plant operation. Note that Capsule Z would follow an equivalent subsequent withdrawal schedule to Capsule V if it is relocated to any 40° capsule location during EOC 21, 22 or 23, since Capsule Z and Capsule V are both currently at 4° effective azimuthal locations and are thus radiologically equivalent.

12

PRESSURE TEMPERATURE LIMITS REPORT Table 5-1 Comparison of the Sequoyah Unit 1 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)

Lower Shell T 0.241 59.85 67.52 16 16 Forging 04 U 0.693 89.3 109.7 20.5 21 (Tangential)

X 1.16 102.6 145.12 23 8 (Heat # 980919 /

281587) Y 1.97 114.95 129.87 26.5 23 Lower Shell T 0.241 59.85 50.59 16 0 Forging 04 U 0.693 89.3 67.59 20.5 19 (Axial)

X 1.16 102.6 103.34 23 22 (Heat # 980919 /

281587) Y 1.97 114.95 133.35 26.5 19 Weld Metal T 0.241 111.13 127.79 35 30 (d)

(Heat # 25295) U 0.693 165.82 144.92 42 26 X 1.16 190.51 159.02 45 21 Y 1.97 213.44 163.8 48 28 HAZ Metal T 0.241 -- 45.48 -- 20 U 0.693 -- 78.94 -- 26 X 1.16 -- 95.89 -- 3 Y 1.97 -- 73.3 -- 10 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 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 REPORT Table 5-2 Calculation of Chemistry Factors using Sequoyah Unit 1 Surveillance Capsule Data Material Capsule Capsule f(a) FF(b) RTNDT(c) FF*RTNDT FF2 Lower Shell T 2.41E+18 0.615 67.52°F 41.52°F 0.378 Forging 04 U 6.93E+18 0.897 0.805 109.70°F 98.42°F (Tangential)

X 1.16E+19 1.041 145.12°F 151.13°F 1.085 (Heat # 980919 /

281587) Y 1.97E+19 1.185 129.87°F 153.92°F 1.405 Lower Shell T 2.41E+18 0.615 50.59°F 31.11°F 0.378 Forging 04 U 6.93E+18 0.897 0.805 67.59°F 60.64°F (Axial)

X 1.16E+19 1.041 103.34°F 107.62°F 1.085 (Heat # 980919 /

281587) Y 1.97E+19 1.185 133.35°F 158.04°F 1.405 SUM: 802.39°F 7.344 CF04 = (FF

• RTNDT) ÷ ( FF2) = (802.39) ÷ (7.344) =109.3°F Surveillance Weld T 2.41E+18 0.615 115.01°F 70.72°F 0.378 Material(d) (127.79°F)

(Heat # 25295)(e) U 6.93E+18 0.897 130.43°F 117.01°F 0.805 (144.92°F)

X 1.16E+19 1.041 143.12°F 149.05°F 1.085 (159.02°F)

Y 1.97E+19 1.185 147.42°F 174.72°F 1.405 (163.80°F)

SUM: 511.50°F 3.672 CF Surv. Weld = (FF

• RTNDT) ÷ ( FF2) = (511.50°F) ÷ (3.672) = 139.3°F Notes:

(a) f = Calculated fluence, (n/cm2, E > 1.0 MeV), updated per TLAA, WCAP-17539-NP, Revision 1[15].

(b) FF = fluence factor = f(0.28 - 0.1*log f) .

(c) RTNDT values are the measured 30 ft-lb shift values taken from Table 5-10 of WCAP-15224[7].

(d) The surveillance weld metal RTNDT 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, Lot #

1103.

14

PRESSURE TEMPERATURE LIMITS REPORT Table 5-3 Reactor Vessel Beltline and Extended Beltline Material Unirradiated Toughness Properties for Sequoyah Unit 1 Fracture Toughness Chemical Composition Properties Initial Material Description Cu Ni P Mn Initial Upper-Shelf (Wt. %) (Wt. %) (Wt. %) (Wt. %) RTNDT(a) (°F) Energy (ft-lb)

Reactor Vessel Beltline Materials(b)

Intermediate Shell Forging 05 0.15 0.86 0.011 0.70 40 79 (Heat # 980807/281489)

Lower Shell Forging 04 0.13 0.76 0.015 0.62 73 72 (Heat # 980919/281587)

Intermediate to Lower Shell Forging Circumferential Weld Seam 0.35 0.11 0.021 1.47 -40 113 W05 (Heat # 25295) (c, e)

Surveillance Weld (d, e) 0.39 0.11 0.021 1.40 --- ---

(Heat # 25295)

Reactor Vessel Extended Beltline Materials Upper Shell Forging 06 0.16 0.89 0.011 0.70 23 83 (Heat # 980950/282758)(f)

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

Upper Shell to Intermediate Shell Circumferential Weld W06 0.17(g) 1.0(g) 0.013(g) 1.90(g) 10(h) 78(h)

(Heat # 25006)

Lower Shell to Bottom Head Ring 0.35 0.11 0.021 1.47 -40 113 Weld W04 (Heat # 25295)(i)

Notes:

(a) The Initial RTNDT values are measured values (b) Except for the best-estimate P and Mn weight percent values, the beltline material properties were taken from WCAP-15293, Revision 2[10]. The weight percent P and Mn values for the beltline forging materials are based on Sequoyah Unit 1 CMTR data. The weight percent P and Mn values for the beltline and surveillance weld materials were determined using Rotterdam weld certification records as well as WCAP-8233[1]

(c) The surveillance weld and the three Rotterdam tests are averaged together for the Best Estimate of the Intermediate to Lower Shell Forging Circumferential Weld Seam Cu wt. % value. The three Rotterdam test results were: 0.30, 0.25 and 0.46 % copper, as referenced from the NRC Reactor Vessel Integrity Database (RVID) and ultimately from Rotterdam Weld Certifications. The Ni wt. % value is identical to the Ni wt. %

value of the surveillance weld.

(d) These copper and nickel values are best estimate values for only the surveillance weld metal and are the average of three data points [0.424 (WCAP-10340, Rev.1), 0.406 (WCAP-10340, Rev.1), 0.33 (WCAP-8233) copper and 0.084 (WCAP-10340, Rev.1), 0.085 (WCAP-10340, Rev.1), 0.17 (WCAP-8233) nickel.]. These values are 15

PRESSURE TEMPERATURE LIMITS REPORT treated as one data point in the calculation of the best estimate average for the inter. to lower shell circ. weld shown above. Originally the 0.424 / 0.406 and 0.084 / 0.085 values were reported as single points, 0.41 - 0.42 and 0.08 (Per WCAP-10340, Rev. 1[7d]), but it is actually made up of two data points. Sample TW58 from Capsule T was broken into two samples, TW58a and TW58b, thus providing the two data points.

(e) Circumferential Weld Seam W05 was fabricated with weld wire type SMIT 40, Heat # 25295, Flux type SMIT 89, Lot # 2275. The surveillance weld was fabricated with weld wire type SMIT 40, Heat # 25295, Flux type SMIT 89, Lot # 1103 and is representative of the intermediate to lower shell circumferential weld.

(f) The chemical compositions for the extended beltline forging materials are based on Sequoyah Unit 1 certified data, except for the weight percent copper values. The maximum weight percent copper value for A508 Class 2 forging materials is conservatively applied, as described in WCAP-17539-NP, Revision 1[15].

(g) Except for the weight percent nickel, the chemical compositions were taken from a chemical analysis performed on the weld wire (heat # 25006) included in the Rotterdam weld certification records. A value of 1.0 was conservatively assumed, as described in WCAP-17539-NP, Revision 1[15].

(h) The initial RTNDT was determined using all available measured data for heat # 25006 as described in WCAP-17539-NP, Revision 1[15]. In absence of USE data for weld heat # 25006, weld heat # 25295 test results from the first surveillance capsule withdrawn from Sequoyah Unit 1 were used, also as described in WCAP-17539-NP, Revision 1[15].

(i) The Lower Shell to Bottom Head Ring Weld was fabricated using the same weld wire heat number and flux type as the Intermediate Shell to Lower Shell Circumferential Weld.

16

PRESSURE TEMPERATURE LIMITS REPORT Table 5-4 Peak Neutron Fluence Projections on the Reactor Vessel Clad/Base Metal Interface for Sequoyah Unit 1 (x 1019 n/cm2, E > 1.0 MeV)

Fluence(a)

Reactor Vessel Material 32 EFPY 52 EFPY Upper Shell Forging 06 0.0369 0.0584 Intermediate Shell Forging 05 1.73 2.66 Lower Shell Forging 04 1.73 2.66 Bottom Head Ring 03 0.215 0.336 Upper Shell to Intermediate Shell 0.0369 0.0584 Circumferential Weld W06 Intermediate Shell to Lower Shell 1.72 2.65 Circumferential Weld W05 Lower Shell to Bottom Head Ring Weld 0.215 0.336 W04 Note:

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

17

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

Material RG 1.99 CF FF IRTNDT(b RTNDT( Margin(d ART(e)

) c) )

R2 Method (°F) (°F)

(°F) (°F) (°F)

Reactor Vessel Beltline Materials Intermediate Shell Forging 05 Position 115.6 1.1301 40 130.6 34.0 204.6 1.1 Position 95.0 1.1301 73 107.4 34.0 214.4 1.1 Lower Shell Forging 04 Position 109.3 1.1301 73 123.5 17.0(f) 213.5 2.1 Intermediate to Lower Shell Position 161.3 1.1291 -40 182.1 56.0 198.1 Circumferential Weld Seam 1.1 Position 139.3 1.1291 -40 157.3 28.0(f) 145.3 2.1 Reactor Vessel Extended Beltline Materials Position 123.9 0.2408 23 29.8 29.8 82.7 Upper Shell Forging 06 1.1 Position 122.3 0.5722 5 70.0 34.0 109.0 Bottom Head Ring 03 1.1 Upper Shell to Intermediate Position 207.0 0.2408 10 49.8 49.8 109.7 Shell Circumferential Weld 1.1 W06 (Heat #25006)

Position 161.3 0.5722 -40 92.3 56.0 108.3 Lower Shell to Bottom Head 1.1 Ring Weld W04 (Heat

1. 25295) Position 139.3 0.5722 -40 79.7 28.0(f) 67.7

2.1 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) RTNDT = CF

• FF (d) Margin = 2 *(i2 + 2)1/2 (e) ART = Initial RTNDT + RTNDT + Margin (°F)

(f) Data deemed credible (See Reference 15), thus the reduced will be used to determine margin for the surveillance forging and surveillance weld material.

18

PRESSURE TEMPERATURE LIMITS REPORT Table 5-6 Sequoyah Unit 1 Calculation of the ART Values for the 3/4T Location at 52 EFPY(a)

Material RG 1.99 CF FF IRTNDT(b RTNDT(c Margin(d ART(e)

) ) )

R2 Method (°F) (°F)

(°F) (°F) (°F)

Reactor Vessel Beltline Materials Intermediate Shell Forging 05 Position 115.6 0.8481 40 98.0 34.0 172.0 1.1 Position 95.0 0.8481 73 80.6 34.0 187.6 1.1 Lower Shell Forging 04 Position 109.3 0.8481 73 92.7 17.0(f) 182.7 2.1 Intermediate to Lower Shell Position 161.3 0.8471 -40 136.6 56.0 152.6 1.1 Circumferential Weld Seam Position 139.3 0.8471 -40 118.0 28.0(f) 106.0 2.1 Reactor Vessel Extended Beltline Materials Position 123.9 0.1291 23 16.0 16.0 55.0 Upper Shell Forging 06 1.1 Position 122.3 0.3579 5 43.8 34.0 82.8 Bottom Head Ring 03 1.1 Upper Shell to Intermediate Position 207.0 0.1291 10 26.7 26.7 63.4 Shell Circumferential Weld 1.1 W06 (Heat #25006)

Position 161.3 0.3579 -40 57.7 56.0 73.7 Lower Shell to Bottom Head 1.1 Ring Weld W04 (Heat

1. 25295) Position 139.3 0.3579 -40 49.9 28.0(f) 37.9

2.1 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) RTNDT = CF

• FF (d) Margin = 2 *(i2 + 2)1/2 (e) ART = Initial RTNDT + RTNDT + Margin (°F)

(f) Data deemed credible (See Reference 15), thus the reduced will be used to determine margin for the surveillance forging and surveillance weld material.

19

PRESSURE TEMPERATURE LIMITS REPORT Table 5-7 Summary of the Sequoyah Unit 1 Reactor Vessel Beltline and Extended Beltline Material ART Values Material RG 1.99 R2 1/4 ART 3/4 ART Method (°F) (°F)

Reactor Vessel Beltline Materials Intermediate Shell Forging 05 Position 1.1 204.6 172.0 Position 1.1 214.4 187.6 Lower Shell Forging 04 Position 2.1 213.5 182.7 Intermediate to Lower Shell Position 1.1 198.1 152.6 Circumferential Weld Seam Position 2.1 145.3 106.0 Reactor Vessel Extended Beltline Materials Upper Shell Forging 06 Position 1.1 82.7 55.0 Bottom Head Ring 03 Position 1.1 109.0 82.8 Upper Shell to Intermediate Position 1.1 109.7 63.4 Shell Circumferential Weld W06 Lower Shell to Bottom Head Position 1.1 108.3 73.7 Ring Weld W04 Position 2.1 67.7 37.9 20

PRESSURE TEMPERATURE LIMITS REPORT Table 5-8 Sequoyah Unit 1 Calculation of the ART Values for the Inlet and Outlet Nozzles at 52 EFPY(a)

Fluence IRTNDT(b) RTNDT Margin(c) ART(d)

Material Heat # (x 1019 n/cm2,

(°F) (°F) (°F) (°F)

E > 1.0 MeV)

Inlet Nozzle 11 4846 0.00358 -15 0 0 -15 Inlet Nozzle 12 4849 0.00358 -1 0 0 -1 Inlet Nozzle 13 4863 0.00358 -22 0 0 -22 Inlet Nozzle 14 4865 0.00358 -31 0 0 -31 Outlet Nozzle 15 4845 0.00358 -19 0 0 -19 Outlet Nozzle 16 4850 0.00358 -10 0 0 -10 Outlet Nozzle 17 4862 0.00358 -24 0 0 -24 Outlet Nozzle 18 4864 0.00358 -40 0 0 -40 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.

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

(d) ART = Initial RTNDT + RTNDT + Margin (°F); RTNDT = 0 for fluence values less than 1 x 1017 n/cm2 (E > 1.0 MeV) per Reference 17.

Table 5-9 Summary of the Sequoyah Unit 1 Nozzle Material ART Values at 52 EFPY Nozzle Material Limiting ART Value (°F)

Inlet Nozzle 12 Heat # 4849 -1 Outlet Nozzle 16 Heat # 4850 -10 21

PRESSURE TEMPERATURE LIMITS REPORT Table 5-10 RTPTS Calculations for Sequoyah Unit 1 Beltline and Extended Beltline Materials at 52 EFPY(a)

Material Fluence FF CF RTPTS(b) Margin RTNDT(U)(c RTPTS(d)

)

(x 1019 n/cm2, (°F) (°F) (°F) (°F)

E>1.0 MeV) (°F)

Reactor Vessel Beltline Materials Intermediate Shell Forging 05 2.66 1.2616 115. 145.8 34.0 40 219.8 6

Lower Shell Forging 04 2.66 1.2616 95.0 119.8 34.0 73 226.8 Lower Shell Forging 04 2.66 1.2616 109. 137.9 17.0(e) 73 227.9 (Using S/C Data) 3 Circumferential Weld Metal 2.65 1.2607 161. 203.3 56.0 -40 219.3 3

Circumferential Weld Metal 2.65 1.2607 139. 175.6 28.0(e) -40 163.6 (Using S/C Data) 3 Reactor Vessel Extended Beltline Materials Upper Shell Forging 06 0.0584 0.3180 123.9 39.4 34.0 23 96.4 Bottom Head Ring 03 0.336 0.6997 122.3 85.6 34.0 5 124.6 Upper Shell to Intermediate Shell Circumferential Weld W06 0.0584 0.3180 207.0 65.8 56.0 10 131.8 Lower Shell to Bottom Head Ring Weld W04 0.336 0.6997 161.3 112.9 56.0 -40 128.9 Lower Shell to Bottom Head Ring Weld W04 0.336 0.6997 139.3 97.5 28.0(e) -40 85.5 (Using S/C Data)

Notes:

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

(b) RTPTS = CF

• FF (c) Initial RTNDT values are measured values (d) RTPTS = RTNDT(U) + RTPTS + Margin (°F)

(e) Data deemed credible (See Reference 15), thus the reduced will be used to determine margin for the surveillance forging and surveillance weld material.

22

PRESSURE TEMPERATURE LIMITS REPORT 6.0 References

1. WCAP-8233, Tennessee Valley Authority Sequoyah Unit No. 1 Reactor Vessel Radiation Surveillance Program, S. E. Yanichko, et. al., December 1973.

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 of Metallic Materials, in ASTM Standards, 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 Failure

5. ASTM E185-82, Annual Book of ASTM Standards, Section 12, Volume 12.02, Standard Practice for 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 Y from the Tennessee Valley Authority Sequoyah Unit 1 Reactor Vessel Radiation Surveillance Program, T.J. Laubham, et. al., June 1999.

7b. WCAP-13333, Analysis of Capsule X from the Tennessee Valley Authority Sequoyah Unit 1 Reactor Vessel Radiation Surveillance Program, M.A. Ramirez, S. L. Anderson, L. Albertin, June 1992.

7c. SwRI Project 06-8851, Reactor Vessel Material Surveillance Program 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 from the Tennessee Valley Authority Sequoyah Unit 1 Reactor Vessel Radiation Surveillance Program, S.E. Yanichko, et. al., February 1984.

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-15293, Revision 2, Sequoyah Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation, J.H. Ledger, July 2003.

11. Westinghouse Letter to TVA, TVA-93-105, Cold Overpressure Mitigation System Code Case and Delta-P Calculation, dated May 19, 1993.

12. Calculation SQN-IC-014, Demonstrated Accuracy Calculation for Cold Overpressure Protection System.

23

PRESSURE TEMPERATURE LIMITS REPORT

13. ASME Code Case N-640, Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1, dated February 26, 1999.

14. WCAP-15984-P, Revision 01, Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Sequoyah Units 1 and 2, W. Bamford, et.al., April 2003.

15. WCAP-17539-NP, Revision 1, Sequoyah Units 1 and 2 Time-Limited Aging Analysis on Reactor Vessel Integrity, E.J. Long, May 2015.

16. MCOE-LTR-16-12-NP, Revision 0, Pressure-Temperature Limit Curve Development for Extended Plant Operation for 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 Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, dated October 14, 2014.

18. LTR-SCS-16-25, Revision 0, Cold Overpressure Mitigation System (COMS) Setpoint Analysis for Sequoyah Units 1 and 2 PORV Replacement, R.A. Freund, B.D. Jaskiewicz, November 2016.

19. LTR-SCS-17-43, Revision 0, Sequoyah LTOPS Setpoint Implementation for PORV Replacement Program Delay, R.A. Freund, September 2017.

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