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

Tennessee Valley Authority, Post Office Box 2000, Soddy Daisy, Tennessee 37384-2000 November 13, 2015 10 CFR 50.4 ATT-N: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Sequoyah Nuclear Plant, Units I and 2 Renewed Facility Operating License Nos. DPR-77 and DPR-79 NRC Docket Nos. 50-327 and 50-328

Subject:

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

References:

1. Letter from NRC to TVA, "Sequoyah Nuclear Plant, Units I and 2

- Issuance of Amendments for the Conversion to the Improved Technical 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 Technical Specifications (Tss) 5.6.4.c, enclosed is the Unit I Pressure Temperature Limits Report (PTLR), Revision 5, and Unit 2 PTLR, Revision 6. In accordance with TSs 5.6.4.c, the PTLRs are required to be provided to the Nuclear Regulatory Commission (NRC) within 30 days after any revision. Sequoyah Units I and 2 were issued license amendment Nos. 334 and 327, respectively for improved standard TSs (Reference 1). These license amendments resulted in the enclosed revisions to each of the PTLRs. The revisions also include other editorial clarifications and administrative changes identified during the revision process as described in . 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 Commission Page 2 November 13, 2015 Sequo Nuclear Plant Enclosures

1. Units 1 and 2 Pressure Temperature Limits Report Changes

2. Sequoyah Unit 1 Pressure Temperature Limits Report, Revision 5

3. Sequoyah Unit 2 Pressure Temperature Limits Report, Revision 6 ZTK: DVG Enclosures cc (Enclosures):

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

ENCLOSURE1I SEQUOYAH UNITS 1 AND 2 PRESSURE TEMPERATURE LIMITS REPORT CHANGES The following describes the editorial clarifications and administrative changes made to each Units' Pressure Temperature Limits Report (PTLR).

1. Conflict between the figure index referencing Cold Overpressure Mitigation System (COMS) and the figure title referencing Low Temperature Overpressure Protection System (LTOPS) was Corrected for consistency with the system terminology in Section 3.4.12 of the Technical Specifications.

2. Section 1, "RCS Pressure Temperature Limits Report (PTLR)," was revised to clarify which Limiting Condition for Operations are affected by the PTLR.

3. Section 3.1, "Pressurizer PORV Lift Setting Limits," was revised for consistency with the analysis of record contained in Topical Report No.

WCAP-1 5293, Revision 2 "Sequoyah Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation," and Topical Report No. WCAP-15321 Revision 2 "Sequoyah Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation".

4. Section 5.0, "Supplemental Data Tables," added the word "Forging" to the sentence in refer to the limiting beltline material identified in Tables 5-5 and 5-6.

5 A note was added to Table 4-1, "Sequoyah Unit 1 Reactor Vessel Surveillance Capsule Withdrawal Schedule," to describe current and future activity with the surveillance capsule withdrawal schedule.

6. Tables 5-5 and 5-6 were revised with a different annotation convention consistent with the other tables in the PTLR.

ENCLOSURE 2 SEQUOYAH UNIT 1 PRESSURE TEMPERATURE LIMITS REPORT, REVISION 5

B88 15 t 16 800 MA PRESSURE TEMPERATURE LIMITS REPORT

.... APPROVED This qp~provaIl does. not reliev, the Contractor from any port o9 his re-,

ponsibility Jar, the correctness of

.* design. debugailnd dimlnsio~ns.

YLetter lo.Nl107l 6 .

oot,: ,October 16, 2015. -.

SQEP i.*% BY: W. JXPierce Tennessee Valley Authority Sequoyah unit 1 Pressure Temperature Limits Report Revision 5, September 2015 PROJECT Sequovah .DISCIPLINE N "CONTRACT .4411 1JNT' 1 DESC. RCS Pressure-Temperature Limit Report

DWG/DOC NO. PTLR-1 SHEET - OF -. "REV. 05

DATE 10/16/15 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 (PIT) Limits (TS 3.4.3) ............................................... 1 3.0 Low Temperature Overpressure Protection System (TS 3.4.12)...... ............................ 1 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................................................................................................ 19 ii

PRESSURE TEMPERATURE LIMITS REPORT List of Tables Table 2-1 Sequoyah Unit 1 Heatup Limits at 32 EPPY (with Uncertainties for Instrumentation Errors of 10°F and 60 psig)......... ...................... 6 Table 2-2 Sequoyah Unit 1 Cooldown Limits at 32 EFPY (with Uncertainties for Instrumentation Errors of 10°F and 60 psig)................................S Table 3-1 Selected Setpoints, Sequoyah Unit 1.......................................................... 10 Table 4-1 Sequoyah Unit I 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 Material Unirradiated Toughness Properties for Sequoyah Unit 1............................................................................ ..... 15 Table 5-4 Peak Neutron Fluence Projections at Key Azimuthal Locations on the Reactor Vessel Clad/Base Metal Interface for Sequoyah Unit 1 (x 10'9 n/cm 2, B> 1.0 MeV) ............ 16 Table 5-5 Sequoyah Unit 1 Calculation of the ART Values for the 1/4T Location @ 32 EFPY.....17 Table 5-6 Sequoyah Unit 1 Calculation of the ART Values for the 3/4T Location @ 32 EFPY ...... 17 Table 5-7 Summary of the Limiting ART Values Used in the Generation of the Sequoyah Unit 1 Heatup/Cooldo~vn Curves. .................................................................... 18 Table 5-8 RTm-s Calculatiohis for Sequoyah Unit 1 Beitline Region Materials at 32 EFPY.......... 18 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 32 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 32 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 Tecirnical 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 (PiT) Limits and TS 3.4.12, Low Temperature Over Pressure Protection (LTOP) System.

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 ofASME 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 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 (PIT) Limits (TS 3.4.3) 2.1.1 The minimum boltup temperature is 50 0 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 l00OF in any one hour period.

c. A maximum temperature change of less than or equal to I10°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.

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 the subsections which follow. These lift setpoinlts have been developed using the NRC-approved methodologies specified in TS 5.6.4.

1

PRESSURE TEMPERATURE LIMITS REPORT 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 32 EFPY steady-state curves (Table 2-2), which are beitline conditions and are not compensated for pressure differences between the pressurizer transmitter and the reactor midplane/beltline or 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 SOTF thermal transport effect for heat injection transients. A demonstrated accuracy calculation (Reference 12) has been performed to confirn 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 Anning 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 r1l) is in compliance with Appendix H to 10 CFR 50, "Reactor Vessel Material Surveillance Program Requirementsr 2 1."' The material test requirements and the acceptance standard utilize the reference nil-ductility temperature RT*T, which is determined in accordance with ASTM E23 [3]. The empirical relationship between RTNDr 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 0, "Fracture Toughness Criteria for Protection Against FailureE41. The surveillance capsule removal schedule meets the requirements ofASTM E185-82N5 . 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 calcuilation 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 32 EFPY for each beitline material in the Sequoyah Unit 1 reactor vessel. The limiting beitline 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 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 REPORT MATERIAL PROPERTY BASIS LTMITING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART VALUES AT 32 EFPY: 1/4T, 216°F 3/4T, 186°F 2500 --_o- erlim Version.5.1 Run.15680 [i /-

2250i Leak Test Limit * /-; -

2000 ....... Unacceptable

,Operation, ....... . .. -- I-* -- "Acceptable Operation ...

__:Heatup Rate *

Critical Limit

_* 100o Deg. F/HrI 100 Deg. F/HrI u*1500 -- "~

2=i (n 'i

. ,.. 120 .. . ...

S 1000 _ __- --- ___ __ _

750 -- -- _ _ -__ -__ ___-__ __ __ ---

i* /

Criticality Limit based on I

inservice hydrostatic test 500 - temnperatuire (288 0F) for the __

service period up to 32 EFPY 250 ~

(IMinimum Boltup .. ... -;- ___

__) _

STemp = 50*F 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 32 EFPY (w/Margins for Instrumentation Error of 10°F and 60 psig) (Plotted Dataprovided on Table 2-1) 4

PRESSURE TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL FORGING 04 LIMITING ART VALUES AT 32 EFPY: 114T, 216°F 3/4T, 186°F 2500 2250 2000 1750

- 1500 21250 o 1000 750 500 250 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 32 EFPY (w/Margins for Instrumentation Error of 10°F and 60 psig) (Plotted Dataprovided on Table 2-2) 5

PRESSURE TEMPERATURE LIMITS REPORT Table 2-1 Sequoyah 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 Limit T P J T Pj 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 2-1 - (Continued)

Sequoyah Unit 1 Heatup Limits at 32 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 32 EFPY (with Uncertainties for Instrumentation Errors of 10°F and 60 psig)

Steady State [20F )40F P1 60F 100F 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

+/- _______________________________________ I 8

PRESSURE TEMPERATURE LIMITS REPORT Table 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 P 220 874 220 853 220 836 220 821 220 806 225 909 225 892 225 878 225 869 2-25 865 230 948 230 935 230 925 230 921 235 991 235 982 235 978 240 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, Sequoyah Unit 1 Trs Dg.) PORV#2 PORV#1 Trcs(De.F)Setpoint (psig) Setpoint (psig) 50 490 465 100 500 475 135 540 510 175 575 540 200 610 570 250 745 685 280 745 685 405 745 685 450 2350 2350 10

PRESSURE TEMPERATURE LIMITS REPORT Sequoyah Unit I LTOPS Selected Setpoints 2500 2000 2

i1500 1000 0

500 0 50 100 150 200 250 300 350 400 450 500 Reactor Coolant System Temnperature (°F)

--4-- FOFRV#2 Setpoint --=- PORV#1 Setpoint Figure 3-1 Sequoyah Unit 1 Selected LTOPS Setpoints (PlottedDataprovided on Table 3-1) 11

PRESSURE TEMPERATURE LIMITS REPORT Table 4-1 Seqtuoyah Unit 1 Reactor Vessel Surveillance Capsule Withdrawal Schedule

Removal Time Fluence Capsule Location Lead Factorca) (EFPY) (b) (n/cm 2 ,E>l.0 MeV)<*1 T 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) Updated in Capsule Y dosimetry analysis (WCAP-15224*71).-

(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 5, V, W and Z will reach a fluence of 2.74 x 1019 (E > 1.0 Mev), the 48 EFPY peak vessel fluence at approximately 44 EFPY, respectively.

Administrative Note - The surveillance capsule withdrawal schedule in Table 4-1 is based on the surveillance program for the original 40 year service life. Relocation of select standby capsules to increase the fluence lead factor in anticipation of the updated surveillance program for the 60 year license renewal 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 to TVA dated September 4, 2015 (ML15244B222). A complete update of the reactor vessel surveillance program for the 60 year license renewal service life will be documented by a subsequent revision to the PTLR prior to entry into the license renewal extended operating period.

12

PRESSURE TEMPERATURE LIMITS REPORT Table 5-1 Comparison of the Sequoyah Unit 1 Surveillance Material 3 0 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/cm 2). (OF)(a) (OF)(b) (%)(a) .(%)(C)

Lower Shell T 0.261 .59.85 67.52 16 16 Forging 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 23 Lower Shell T 0.261 59.85 50.59 16 0 Forging 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 19 Weld Metal T 0.261 111.13 127.79 . 35 30 (Heat # 2 52 95 )(d) U 0.796 165.82 144.92 42 26 X 1.32 190.51 159.02 . 45 .21 Y 2.19 213.44 163.8 48 28 IiAZ Metal T 0.261 -- 45.48 -- 20 U 0.796 - -. 78.94 -- 26 X 1.32 -- 95.89 -- 3 Y 2.19 -- 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) ARTNBT(C) FF*ARTNDT FF2 Lower Shell T 2.61E+18 0.63 67.52°F 42.54°F 0.40 Forging 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 1.47 Lower Shell T 2.6 1E+I18 0.63 50.59°F 31.87°F 0.40 Forging 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 1.47 281587)

SUM: 827.89°F 7.82 CF04 = Y*(FF 2

RTNDTr) + 2( FF ) = (827.89) +(7.82) = 105.9°F Surveillance Weld T 2.61E+18 0.63 115.0°F 72.5°F 0.40 Material(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 1.16 Y 2.19E+19 1.21 147.4OF 178.4 0 F 1.47 SUM: 528.0°F 3.91 CF Surv.we~d =XY(FF

RTrrDT) +X( FF2) =(528.0°F) + (3.91) = 135.0°F Notes:

(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 one decimal 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 1103 14

PRESSURE TEMPERATURE LIMITS REPORT Table 5-3 Reactor Vessel Beltline Material Unirradiated Toughness Properties for Sequoyah Unit 1 Material Description Cu (%) Ni (%) Initial RTNDT(a)

Intermediate Shell Forging 05 0

(Heat'#980807/281489) 01 .64 Lower Shell Forging 04 (Heat #980919/281587) 0 01 .67 Surveillance Weld (Heat # 2 5 2 9 5 )(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 Shell Forging Circumferential Weld Seam W05 0.35 0.11 -40OF (Heat # 2 5 2 9 5 )(d. e)

N*otes:

(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 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 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. l[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.

(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 SMVIT 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.

(e) 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.

15

PRESSURE TEMPERATURE LIMITS REPORT Table 5-4 Peak Neutron Fluence Projections at Key Azimuthal Locations on the Reactor Vessel Clad/Base Metal Interface for Sequoyah Unit 1 (x 10 9n/cm 2, E > 1.0 MeV)

Azimuthal Location EFPY 00 150 300 450 10.03 0.205 0.321 0.409 0.637 20 0.387 0.596 0.761 1.18 32 0.605 0.928 1.19 1.84 48 0.896 1.37 1.75 2.72 16

PRESSURE TEMPERATURE LIMITS REPORT Table 5-5 Sequoyah Unit 1 Calculation of the ART Values for the 1/4T Location @ 32 EFPY(a)

Material RG 1:99 CF FF IRT*,T(b) ARTnDT(C) Margin(d) ART(e)

R2 Method (0F) (°F) (0F) (0 F) (0F)

Intermediate Shell Forging 05 Position 1.1 115.6 1.029 40 119.0 34 193 Position 1.1 95 1.029 73 97.8 34 205 Lower Shell Forging 04 Position 2.1 105.9 1.029 73 109.0 34(0 216 Intermediate to Lower Shell Position 1.1 161.3 1.029 -40 166.0 56 182 Circumferential Weld Seam Position 2.1 135.0 1.029 -40 138.9 56(0 155 Notes:

(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 5-6 Sequoyah 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) 0 0 0 0

R2 Method ( F) ( F) ( F) ( F) (0F)

Intermediate Shell Forging 05 Position 1.1 115.6 0.747 40 86.4 34 160 Loe 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(0 186 Intermediate to Lower Shell Position 1.1 161.3 0.747 -40 120.5 56 137 Circumferential Weld Seam Position 2.1

135.0 0.747 -40 100.8 56(0 117 Notes:

(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 GrAwill be used to determine margin.

17

PRESSURE TEMPERATURE LIMITS REPORT Table 5-7 Summnary of the Sequoyah Unit 1 Reactor Vessel Beitline Material ART Values Material RG 1.99 R2 1/4 ART 3/4 ART Method (0 F) (0 F)

Intermediate Shell Forging 05 Position 1.1 193 160 Position 1.1 205 178 Lower Shell Forging 04 Position 2.1 216 186 Intermediate to Lower Shell Position 1.1 182 137 Circumferential Weld Seam Position 2.1 155 117 Table 5-8 RTP-s Calculations for Sequoyah Unit 1 Beltline .Region Materials at 32 EFPY(a)

Material Fluence FF CF ARTpTs~b) Margin RTrNDT*(U)) RTpxs(d)

(X 1019 n/cm 2 , "(F) ( 0 F) (0 F) (OF) ( 0 F)

E>1.0 MeV)

Intermediate Shell Forging 05 1.84 1.167 115.6 .134.9 34 40 209 Lower Shell Forging 04 1.84 1.167 95.0 110.9 34 73 218 Lower 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 -40 204 Circumferential 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= RThrT*u + ARTpTs + Margin (0F)

(e) Data deemed not-credible (See Reference 7a), thus the full *a will be used to determine margin.

18

PRESSURE TEMPERATURE LIMITS REPORT 6.0 References

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

2. Code of Federal Regulations, 10OCFR50, Appendix H, Reactor Vessel MaterialSurveillance Programn Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C..

3. ASTM E23 StandardTest Method Notched Bar himpact 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" Criteriafor ProtectionAgainst Failure

5. ASTM E185-82, Annual Book of ASTM Standards, Section 12, Volume 12.02, StandardPractice for Conducting Surveillance Testsfor 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 Yfr*om the Tennessee Valley Author"ity Sequoyah Unit 1 Reactor Vessel Radiation Surveillance Program, T.J. Laubham, et. al., June 1999.

7b. WCAP-1 3333, Analysis of CapsuleXfoino the Tennessee Valley Authority Sequoyah Unit 1 Reactor Vessel Radiation Surveillance Progr~am, M.A. Ramirez, S. L. Anderson, L. Albertin, June 1992.

7c. .SwRI Project 06-8851, Reactor Vessel MaterialSurveillance Progr'amfor Sequoyah Unit No. 1:

Analysis of Capsule U, P. K. Nair, et al., October 1986.

7d. WCAP- 10340, Revision 1, Analysis of Capsule Tfoin the Tennessee Valley Authority Sequoyah Unit.] 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 OverpressureMitigating 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-01 4, DemonstratedAccuracy Calculationfor Cold Overpressure Protection System.

19

PRESSURE TEMPERATURE LIMITS REPORT

13. ASME Code Case N-640, Alternative Reference Fracture Toughnessfor Development of P-T Limit Curvesfor Section XL, Division 1, dated February 26, 1999.

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

20

ENCLOSURE 3 SEQUOYAH UNIT 2 PRESSURE TEMPERATURE LIMITS REPORT, REVISION 6

Bi88 151.016" 801 PRESSURE TEMPERATURE LIMITs REPORT IContractrfo ny port of his re-.

dligdealsad dinmsnsions. .*

o"*t_. 'Octobe~r 16,.2015. _

SOEP (Nt) BYW. J.Pierce__

Tennessee Valley Authority Sequoyah unit 2 Pressure Temperature Limits Report Revision 6, september 2015 PROJECT Secjuovah 'DisCIPLINE N CONTRACT " 4411 UNIT. 2 DESC. RCS Pressure-Temperature L'mait-Report DWG/DOC NO. PTLR-2 SHEET ..- OF - REV. 06*

DATE. 10/16/15 ECN/DCN FILE N2N-081" EDMS, WT CA-K

PRESSURE TEMPERATURE LIMITS REPORT Table of Contents Litoof.............Tables...........

T be .List. ................. ....................... i List of Figures ............................................................................................. ........ v 1.0 RCS Pressure Temperature Limits RePort (PTLR) ........ .... .................................... 1 2.0 Operating Limits ............................................................................. ............. 1 2.1 RCS Pressure/Temperature (PIT) Limits (TS 3.4.3)................................................. 1 3.0 Low Temperature Overpressure Protection System (TS 3.4.12) ...................... .. 1 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..3 6.0 References ......................................... ..... .. ... *.................................. ........ 17 ii

PRESSURE TEMPERATURE LIMITS REPORT List of Tables Table 2-1 Sequoyah Unit 2 Heatup Limits at 32 BEFPY (with Uncertainties for Instrumentation Errors of l0°F and 60 psig)............................... 6 Table 2-2 Sequoyah Unit 2 Cooldown Limits at 32 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 Material Unirradiated Toughness Properties for Sequoyah Unit2 ................................................................................ 13 Table 5-4 Peak Neutron Pluence Projections at Key Azimuthal Locations on the Reactor Vessel Clad/Base Metal Interface for Sequoyah Unit 2 (x 1019 n/cm 2 , EB> 1.0 MeV) .... i........ 14 Table 5-5 Sequoyah Unit 2 Calculation of the ART Values for the 1/4T Location @ 32 EFPY.....15 Table 5-6 Sequoyah Unit 2 Calculation of the ART Values for the 3/4T Location @ 32 EPPY.....15 Table 5-7 Summary of the Limiting ART Values Used in the Generation of the Sequoyah Unit 2 Heatup/Cooldown Curves.......................................................... :............ 16 Table 5-8 RTPTs Calculations for Sequoyah Unit 2 Beltline Region Materials at 32 EFPY.......... 16 iii

PRESSURE TEMPERATURE LIMITS REPORT List of Figures Figure 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 100 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 First 32 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 Sequoyah Unit 2 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 and TS 3.4.12,.Low Temperature Overpressure Protection (LTOP) System.

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-64011 n1 (Use of K1 o),

WCAP- 159 84-Ptl21 (Elimination of the Flange Requirement), 1996 Version of Appendix G[41 and the revised fluencesETI. 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.

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.

I

PRESSURE TEMPERATURE LIMITS REPORT

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 32 EFPY steady-state curves (Table 2-2), which are beitline conditions and are not compensated for pressure differences between the pressurizer transmitter and the reactor midplane/beltline or 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 set-points 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 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 transmhitter and reactor midplane and maximum temperature/pressure instrument uncertainties are applied to the setpoints.

3.2 Arming Temperature The LTOPS arning temperature is based upon the methodology defined in the Sequoyah Nuclear Plant Unit 2 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-85 13[1]) is in compliance with Appendix H to 10 CFR 50, "Reactor Vessel Material Surveillance Program Requirementst 2 1."' 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 RTNOT 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 Failurer 4k.' The surveillance capsule removal schedule meets the requirements of ASTM E185-82N5 . 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 shaows 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 32 EFPY for each 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 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 REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL! INTERMEDIATE SHELL FORGING 05 LIMITING ART VALUES AT 32 EFPY: 1/4T, 142°F 3/4T, l15°F 2Rflfl Io

)perlim

-v Version:5.1 Run:5694 ILeak, Test, Limit

/

2250 2000- Unacceptable1Oeain ....

-[

Acceptable Operation ...

1750 - * * ..

1500 125

100 ~ritical Deg.Limit*

F/Hr U) 00 Co 0 25 750" Criticality Limit based on 500

-~inservic e hydrostatic test

temperature (214°F) for the service period up to 32 EFPY Boltup Temp 250 -- = 50 0F Minumum "

U . . . . . . .

f I . . . . . . . . . . . . . . ! . . . . ! . . . . . . .. . . . . . . . . .

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

Figure 2-1 Sequoyah Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 100 0 F/hr) Applicable for the First 32 EFFY (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 32 EFPY: 1/4T, 142°F 3/4T, 150 F.

2500 [Oeri Version:5.1 Run:5694]

2000 tUnacceptable _____ --

__ _________ Acceptable Operation OPeration * -

1750 . . ... '

S 1250 * -.- o----__ __ - __ ____ '--

"o Cooldown

~~Rates i

"= 1000 F/Hr .......................

u steady-state o-6o1 750 60oo 250 -.... ........... M inim um I- ...... :.. . ....... .. ... ... 4.. ... . ........

SBoltup Templi . . ! ' i

- 50 0 F !

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

Figure 2-2 Sequoyahi~nit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 32 EFPY (w/ Margins for Instrumentation Error of 10 0F and 60 psig) (PlottedData provided on Table 2-2) 2

PRESSURE TEMPERATURE LIMITS REPORT Table 2-1 Sequoyah 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 Limit T P T P JT 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 6.46 214 671 105 646 214 680 110 646 214 685 115 646 214 701 120 646 214 720 125 648 214 743 130 653 214 769 135 661 214 798 140 671 215 832 145 685 220 869 150 701 225 911 155 720 230 959 160. 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 3

PRESSURE TEMPERATURE LIMITS REPORT Table 2-2 Sequoyah Unit 2 Cooldown Limits at 32 EFPY (with Uncertainties for Instrumentation Errors of 10°F and 60 psig)

Steady State T P 1 20F T P 40F T P 60F T P I00OF 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 176 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 4

PRESSURE TEMPERATURE LIMITS REPORT Table 3-1 Selected Setpoints, Sequoyah Unit 2 Trcs (Deg.F) PORV#2 Setpoint PORV#1 (psig) Setpoint (psig) 50 510 485 100 580 555 135 640 610 174 745 682 200 745 685

250 745 *685-278 745 685 400 745 685 450 2350 2350 5

PRESSURE TEMPERATURE LIMITS REPORT Sequoyah Unit 2 LTOPS Selected Setpoints

"*25000 -/! [

4-2

500 - - J L . . . .

o- I I !

0 01015 0 5 303040 5 0 Reacto Coln Iyteenpraue(F A F

-- *-- FORV#2-FinaI -- U-- FORV#1 Figure 3-1 Sequoyah Unit 2 Selected LTOPS Setpoints (PlottedDataprovided on Table 3-1) 6

PRESSURE TEMPERATURE LIMITS REPORT Table 4-1 Sequoyah Unit 2 Reactor Vessel Surveillance Capsule Withdrawal Schedule (a) Updated in Capsule Y dosimetry analysis (WCAP-15320t 7 1).

(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 peak vessel fluence at approximately 44 EFPY.,

Administrative Note - The surveillance capsule withdrawal schedule in Table 4-1 is based on the surveillance program for the original 40 year service life. Relocation of select standby, capsules to increase the fluence lead factor in anticipation of the updated surveillance program for the 60 year license renewal service life is described in a TVA letter to NRC dated January 10, 2013 (ML13032A251). .

Regulatory approval for the anticipatory standby capsule relocation has been granted by a NRC letter to TVA dated September 27, 2013 (ML13240A320). A complete update of the reactor vessel surveillance program for the 60 year license renewal service life will be documented by a subsequent revision to the PTLR prior to entry into the license renewal extended operating period.

7

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/cm 2 ) (OF)(a) (OF)(b) (%,/)(al) (%)(c)

Intermediate Shell T 0.261 60.33. 63.65 17 12 Forging 05 U 0.692 85.22 79.31 21 16 (Tangential)

(et285/X 1.22 100.23 85.7 23 8 981057) Y 2.14 114.67 134.12 26 22 Intermediate Shell T 0.261 60.33 48.73 17 7 Forging 05 U 0.692 85.22 66.06 21 9 (Axial)

(et285/X 1.22 100.23 110.04 23 2 981057) Y 2.14 . 114.67 89.21 26 22 Weld Metal T 0.261 43.12 74.56 20 2 (Heat # 4 2 7 8 )(d) U 0.692 60.91 130.38 25 6 X 1.22 71.63 44.22 29 35 Y 2.14 81.96 86.91 33 "3 HAZ Metal T 0.261 -- 24.58 -,- 2 U 0.692 - - 64.03 - - 14 X 1.22 - - 28.29 - - 19 Y 2.14 - - 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 [.]

(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 REPORT Table 5-2 Calculation of Chemistry Factors using Sequoyah Unit 2 Surveillance Capsule Data Material Capsule Capsule f(a) FF(b) ARTNDT(C) FF*ARTNDT FF2 Intermediate Shell T 2.61E+18 0.635 63.7 40.45 0.403 Forging 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 1.457 981057)

Intermediate Shell T 2.61E+18 0.635 48.7 30.92 0.403 Forging 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 1.457 981057) ______ ______

SUM: 677.77°F 7.556 CFo5 = X(FF

RTNDT) + X.( FF2 ) = (677.77) +(7.556) = 89.7 0 F Surveillance Weld T 2.6 1E+18 0.635 69.4 (74.6) 44.07 0.403 Material(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 1.113 Y 2.14E+19 1.207 80.8 (86.9) 97.53 1.457 SUM: 293 .77°F 3.778 CF Surv. weld = *(FF

RTNDT) + X( FF2) = (293.77°F) +(3.778) = 77.8"F Notes:

(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 one decimal 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 REPORT Table 5-3 Reactor Vessel Beltline Material Unirradiated Touglmess Properties for Sequoyah Unit 2 Material Description Cu (%) Ni (%) Initial RTNDT(a)

Intermediate Shell Forging 05 (Heat #288757 /981057)013.710F Lower Shell Forging 04 0.14 0.76 -22 °F (Heat # 990469 / 293323)

Intermediate to Lower Shell Forging Circumferential Weld Seam W05(b) 0.12 0.11 -40 F (Heat # 4278)

Surveillance Weld(b) 0.13 0.11 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 of the intermediate to lower shell circumferential weld.

10

PRESSURE TEMPERATURE LIMITS REPORT Table 5-4 Peak Neutron Fluence Projections at Key Azimuthal Locations on the Reactor Vessel Clad/Base Metal Interface for Sequoyah Unit 2 (x 1019n/cm 2, E > 1.0 MeV)

Azimuthal Location EFPY 0° 150 3 00 450 10.54 0.211 0.336 0.426 0.637 20 0.38 0.60 0.773 1.16 32 0.593 0.934 1.21 1.82 48 0.878 1.38 1.80 2.71 11

PRESSURE TEMPERATURE LIMITS REPORT Table 5-5 Sequoyah 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) (0 F) (0 F) (0F)

Position 1.1 95 1.027 10 97.6 34 142 Intermediate Shell Forging 05 Position 2.1 89.7 1.027 10 92.1 34 136 Lower Shell Forging 04 Position 1.1 104 1.027 -22 106.8 34(0 119 Intermediate to Lower Shell " Position 1.1 .63 '1.027 -4 64.7 56 117 Circumferential Weld Seam Position 2.1 77.8 " 1.027 . -4 79.9 56(0 132 Notes:

(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 (0 F)

(f) Data deemed not-credible (See Reference 7a), thus the full ca will be used to determine margin.

Table 5-6 Sequoyah Unit 2 Calculation of the ART Values for the 3/4T Location @ 32 EFPY(a)

Material RG 1.99 CF FF IRTNDT(b) A*TNDT(c) Margin(d) ART(e)

R2 Method (0 F) (0 F) (0 F) (0 F) (0 F)

Position 1.1 95 0.745 10 70.8 34

115 Intermediate Shell Forging 05 Position 2.1 89.7 0.745 10 66.8 34 .111 Lower Shell Forging 04 Position 1.1 104 0.745 -22 77.5 34(0 90 intermediate to Lower Shell Position 1.1 63 0.745 .- 4 46.9 56 99 Circumferential Weld Seam Position 2.1 77.8 0.745 -4 58.0 56(0 110 Notes: .

(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) ART*T =CF *FF (d) M = 2 *(0i2 + raA)2 (e) ART =Initial RTrNDv + ARTNDT + Margin (0 F)

(f) Data deemed not-credible (See Reference 7a), thus the full GA will be used to determine margin.

12

PRESSURE TEMPERATURE LIMITS REPORT Table 5-7 Summary of the Sequoyah Unit 2 Reactor Vessel Beitline Material ART Values Material RG 1.99 R2 1/4 ART 3/4 ART Method (0 F) (0 F)

Position 1.1 142 115 Intermediate Shell Forging 05 Position 2.1 136 111 Lower Shell Forging 04 Position 1.1 119 90 Intermediate to Lower Shell Position 1.1 117 99 Circumferential Weld Seam Position 2.1I 132 110 Table 5-8 RTPTs Calculations for Sequoyah Unit 2 Beltline Region Materials at 32 EFPY(a)

Material Fluence FF CF ARTpTs~b) Margin RTNDT(U-)(C) RTpTs(d)

(x 10 19 n/cm 2, (0F) (0F) (0F) (0F) (0F)

E>l.0 MeV)

Intermediate Shell Forging 05 1.82 1.164 95 110.6 34 10 155 Intermediate 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) -22 133 Circumferential Weld Metal 1.82 1.164 63 73.3 56 -4 125 Circumferential 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 ( 0 F)

(e) Data deemed not-credible (See Reference 7a), thus the full ra..will be used to determine margin.

13

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, 10OCFR50, Appendix H, Reactor Vessel MaterialSurveillance Progr~am Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C.

3. ASTM E23 StandardTest 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 Criteriafor ProtectionAgainst Failure

5. ASTM E185-82, Annual Book ofASTM Standards, Section 12, Volume 12.02, StandardPractice 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- 15320, Analysis of Capsule Y fi'omn the Tennessee Valley Authority Sequoyah Unit 2 Reactor Vessel RadiationSurveillance Program,T.J. Laubham, et. al., November 1999.

7b. WCAP-*10509, Analysis of Capsule Tfiom the. Tennessee Valley Authority Sequoyah Unit 2 Reactor Vessel RadiationSurveillance 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 Programfor 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 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 OverpressureMitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, J.D. Andrachek, et. al., January 1996.

10. WCAP-15321, Revision 2, Sequoyah Unit 2 Heatup and CooldownLimit Curvesfor Normal Operation and PlTLR Support Documentation, J.H. Ledger, et.al., July 2003.

11. ASME Code Case N-640, Alternative Reference FractureToughness for Development of P-T

LimitCurves for Section XL, Division 1, dated February 26, 1999.

14

PRESSURE TEMPERATURE LIMITS REPORT

12. WCAP-15984-P, Revision 01, Reactor Vessel Closure Head/Vessel Flange Requiremnents EvaluationFor Sequoyah Units 1 and 2, W. Bamford, et.al., April 2003.

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

14. Calculation SQN-IC-0 14, DemonstratedAccuracy Calculationfor Cold OverpressureProtection System.