WBL-21-056, Revised Pressure and Temperature Limits Report (PTLR)

Revised Pressure and Temperature Limits Report (PTLR)

ML21343A151
Person / Time
Site:	Watts Bar
Issue date:	12/09/2021
From:	Anthony Williams Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
WBL-21-056
Download: ML21343A151 (32)

Text

U.S. Nuclear Regulatory Commission WBL-21-056 Page 2 December 9, 2021 Enclosure Watts Bar Nuclear Plant, Unit 1 Pressure and Temperature Limits Report (PTLR),

Revision 13.

cc: (Enclosure)

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Watts Bar Nuclear Plant NRC Project Manager - Watts Bar Nuclear Plant

WBL-21-056 E1 of 30 Enclosure Watts Bar Nuclear Plant, Unit 1 Pressure and Temperature Limits Report (PTLR),

Revision 13

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 200 of 255 Appendix A (Page 1 of 29)

Watts Bar Unit 1 - RCS Pressure and Temperature Limits Report (PTLR) - Revision 13 APPENDIX A TO RCS SYSTEM DESCRIPTION N3-68-4001 WATTS BAR UNIT 1 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

REVISION 13 Prepared by:

C. S. Kerlin Checked by S. E. Nolan Approved by:

M. R. Smith WBL-21-056 E2 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 201 of 255 Appendix A (Page 2 of 29) 1.0 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

This PTLR for Watts Bar Unit 1 has been prepared in accordance with the requirements of Technical Specification 5.9.6. Revisions to the PTLR shall be provided to the NRC after issuance.

The Technical Specifications affected by this report are listed below:

LCO 3.4.3 RCS Pressure and Temperature (P/T) Limits LCO 3.4.12 Cold Overpressure Mitigation System (COMS) 2.0 RCS PRESSURE AND TEMPERATURE LIMITS The limits for LCO 3.4.3 are presented in the subsection which follows. These limits have been developed (Ref. 1) using the NRC-approved methodologies (Ref. 4) that are specified in Technical Specification 5.9.6.

2.1 RCS Pressure and Temperature (P/T) Limits (LCO 3.4.3) 2.1.1 The RCS temperature rate-of-change limits are (Ref. 1):

A.

A maximum heatup Rate 100F per hour.

B.

A maximum cooldown Rate 100F per hour.

C.

A maximum temperature change of 10F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

2.1.2 RCS P/T Limits for Heatup, Cooldown, Inservice Hydrostatic and Leak Testing, and Criticality The RCS P/T limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticality are specified by Figures 2.1-1 thru 2.1-3 (Ref. 1).

NOTE:

The heat-up and cool-down curves are based on beltline conditions and do not compensate for pressure differences between the pressure transmitter and reactor midplane/beltline or for instrument inaccuracies. Refer to Table 2.1-3 for pressure differences (Ref. 9). Site Engineering Setpoint and Scaling documents SSD-1-P-68, -63, -64, -66, and -70 provide the adjusted curves for temperature and pressure limits which are compensated for pressure differential and instrument inaccuracy to be used for heatup and cooldown.

NOTE:

Steady-state conditions (0F per hour heatup or cooldown curves) are achieved after maintaining a constant temperature for a duration of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (Ref. 1). Steady state conditions are maintained if temperature fluctuations remain within 9.2F (Ref. 1).

WBL-21-056 E3 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 202 of 255 Appendix A (Page 3 of 29) 3.0 COLD OVERPRESSURE MITIGATION SYSTEM (LCO 3.4.12)

The lift setpoints for the pressurizer Power Operated Relief Valves (PORVs) are presented in the subsection which follows. These lift setpoints have been developed using the NRC-approved methodologies specified in Technical Specification 5.9.6.

3.1 Pressurizer PORV Lift Setting Limits The pressurizer PORV lift setpoints in Table 3.1-1 were specified by the Westinghouse Cold Overpressure Mitigation Setpoint Analysis (Ref. 9). Per ASME Code N-514, the limits for the COMS setpoints are based on 110% of the Steady State 32 EFPY curves (Table 2.1-2)

(Ref. 1) which are based on beltline conditions and are not compensated for pressure differences between the pressure transmitter and the reactor midplane/beltline or for instrument inaccuracies. Refer to Table 2.1-3 for pressure differences (Ref. 9).

NOTE:

These setpoints include allowance for pressure difference between the pressure transmitter and reactor midplane, and also includes 63 psig pressure channel uncertainty. Site Engineering Setpoint and Scaling documents for instrument loop numbers 1-T-68-1B and 1-T-68-43B (Ref. 10, 11) contain the adjusted curves compensated for pressure differential and instrument inaccuracy which provides the PORV lift limits for the COMS utilizing the 32 EFPY data.

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 the most recent examination (Ref. 8) was used to update Figures 2.1-1 through 2.1-3 and the corresponding data points in Table 2.1-1 through 2.1-2 as well as the supplemental data tables in Section 5.0.

The pressure vessel steel surveillance program (Ref. 3) is in compliance with Appendix H to 10 CFR 50, entitled "Reactor Vessel Material Surveillance Program Requirements". The material test requirements and the acceptance standard utilize the reference nil-ductility temperature, RTNDT, which is determined in accordance with ASTM E23. The empirical relationship between RTNDT and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G, "Fracture Toughness Criteria for Protection Against Failure", to Section XI of the ASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM E185-82.

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, predictions.

Table 5-2 shows calculations of the surveillance material chemistry factors using surveillance capsule data.

WBL-21-056 E4 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 203 of 255 Appendix A (Page 4 of 29) 5.0 SUPPLEMENTAL DATA TABLES (continued)

Table 5-3 provides the required Watts Bar Unit 1 reactor vessel toughness data, along with the unirradiated & best estimate average chemistry, respectively. The bolt-up temperature is also included in this table.

Table 5-4 provides a summary of the fluence values used in the generation of the heatup and cooldown limit curves and in the Pressurized Thermal Shock (PTS) evaluation.

Table 5-5 provides a summary of the adjusted reference temperature (ART) values of the Watts Bar Unit 1 reactor vessel beltline materials at the 1/4-T and 3/4-T locations for 32 EFPY.

Table 5-6 shows example calculations of the adjusted reference temperature (ART) values at 32 EFPY for the limiting Watts Bar Unit 1 reactor vessel material (Intermediate Shell Forging 05).

Table 5-7 provides RTPTS values for Watts Bar Unit 1 for 32 EFPY.

6.0 REFERENCES

1.

WCAP-16761-NP, Revision 0, Watts Bar Unit 1 Heatup and Cooldown Curves for Normal Operation, November 2007.

2.

Deleted.

3.

WCAP-9298, Revision 1, "Watts Bar Unit 1 Reactor Vessel Radiation Surveillance Program," April 1993.

4.

WCAP-14040, Revision 1, Methodology Used To Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, December 1994.

5.

Deleted.

6.

BWXT SERVICES, Inc., Analysis of Capsule W From The Tennessee Valley Authority Watts Bar Unit 1 Reactor Vessel Material Surveillance Program, September 2001.

7.

Deleted.

8.

WCAP-16760-NP, Revision 0, Analysis of Capsule Z from the Tennessee Valley Authority, Watts Bar Unit 1 Reactor Vessel Radiation Surveillance Program2, November 2007.

9.

LTR-SCS-15-31, Rev 0, Watts Bar Unit 1 PORV Replacement - Cold Overpressure Mitigation System (COMS) Setpoint Analysis, 1/4/2017 10.

NE SSD 1-T-68-1B, COMS Setpoint and Scaling Document for PORV 1-PCV-68-340A Instrument Loop Components.

WBL-21-056 E5 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 204 of 255 Appendix A (Page 5 of 29)

6.0 REFERENCES

(continued) 11.

NE SSD 1-T-68-43B, COMS Setpoint and Scaling Document for PORV 1-PCV-68-334 Instrument Loop Components.

12.

Structural Integrity Associates Report, No. SIR 01 140, Revision 1, Heatup and Cooldown Limit Curves for Normal Operation for Watts Bar Unit 1, July 2002.

13 MRP-326, Revision 1, Coordinated PWR Reactor Vessel Surveillance Program (CRVSP) June 2021.

WBL-21-056 E6 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 205 of 255 Appendix A (Page 6 of 29) 7.0 FIGURES AND TABLES MATERIAL PROPERTY BASIS LIMITING MATERIAL:

INTERMEDIATE SHELL FORGING 05 LIMITING ART AT 32 EFPY:

1/4-T, 205.74 F 3/4-T, 171.15 F Figure 2.1-1:

Watts Bar Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 60F/hr) Applicable for 32 EFPY (with the Flange-Notch & without Margins for Instrumentation Errors) Using 1996 App. G Methodology (w/KIA)

(Plotted Data (Ref. 1) provided on Table 2.1-1) 0 250 500 750 1000 1250 1500 1750 2000 2250 2500 0

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

Calculated Pressure (PSIG)

Operlim Version:5.2 Run:28921 Operlim.xls Version: 5.2 Unacceptable Operation Acceptable Operation Criticality Limit based on inservice hydrostatic test temperature (331 F) for the service period up to 32 EFPY Heatup Rate 60 Deg. F/Hr Critical Limit 60 Deg. F/Hr Leak Test Limit Boltup Temperature 60 Deg. F WBL-21-056 E7 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 206 of 255 Appendix A (Page 7 of 29) 7.0 FIGURES AND TABLES (continued)

MATERIAL PROPERTY BASIS LIMITING MATERIAL:

INTERMEDIATE SHELL FORGING 05 LIMITING ART AT 32 EFPY:

1/4-T, 205.74 F 3/4-T, 171.15 F Figure 2.1-2:

Watts Bar Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 100F/hr) Applicable for 32 EFPY (with the Flange-Notch & without Margins for Instrumentation Errors) Using 1996 App. G Methodology (w/KIA)

(Plotted Data (Ref. 1) provided on Table 2.1-1) 0 250 500 750 1000 1250 1500 1750 2000 2250 2500 0

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

Calculated Pressure (PSIG)

Operlim Version:5.2 Run:28921 Operlim.xls Version: 5.2 Unacceptable Operation Acceptable Operation Criticality Limit based on inservice hydrostatic test temperature (331 F) for the service period up to 32 EFPY Heatup Rate 100 Deg. F/Hr Critical Limit 100 Deg. F/Hr Leak Test Limit Boltup Temperature 60 Deg. F WBL-21-056 E8 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 207 of 255 Appendix A (Page 8 of 29) 7.0 FIGURES AND TABLES (continued)

Table 2.1-1:

Watts Bar Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate up to 100F/hr) Applicable for 32 EFPY (with the Flange-Notch & without Margins for Instrumentation Errors) Using 1996 App. G Methodology (w/KIA)

(Data points plotted on Figures 2.1-1 and 2.1-2) 60F/hr Heatup Critical Limit 100F/hr Heatup Critical Limit T

[F]

P

[psig]

T

[F]

P

[psig]

T

[F]

P

[psig]

T

[F]

P

[psig]

60 0

331 0

60 0

331 0

60 465 331 465 60 413 331 413 65 465 331 465 65 413 331 413 70 465 331 466 70 413 331 414 75 465 331 467 75 413 331 414 80 465 331 467 80 413 331 416 85 465 331 470 85 413 331 416 90 465 331 470 90 413 331 418 95 465 331 473 95 413 331 419 100 465 331 474 100 413 331 421 105 465 331 477 105 413 331 422 110 466 331 481 110 413 331 425 115 467 331 481 115 413 331 427 120 470 331 487 120 413 331 429 125 473 331 490 125 413 331 433 130 477 331 492 130 413 331 435 135 481 331 499 135 413 331 440 140 487 331 501 140 414 331 441 145 492 331 506 145 416 331 447 WBL-21-056 E9 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 208 of 255 Appendix A (Page 9 of 29) 7.0 FIGURES AND TABLES (continued)

Table 2.1-1:

Watts Bar Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate up to 100F/hr) Applicable for 32 EFPY (with the Flange-Notch & without Margins for Instrumentation Errors) Using 1996 App. G Methodology (w/KIA)

(Data points plotted on Figures 2.1-1 and 2.1-2) 60F/hr Heatup Critical Limit 100F/hr Heatup Critical Limit T

[F]

P

[psig]

T

[F]

P

[psig]

T

[F]

P

[psig]

T

[F]

P

[psig]

150 499 331 514 150 418 331 449 155 506 331 516 155 421 331 455 160 514 331 523 160 425 331 459 165 523 331 533 165 429 331 463 170 533 331 543 170 435 331 471 175 543 331 554 175 441 331 473 180 554 331 566 180 447 331 483 185 566 331 579 185 455 331 486 190 579 331 593 190 463 331 494 195 593 331 608 195 473 331 502 200 608 331 625 200 483 331 506 205 625 331 642 205 494 331 519 210 642 331 661 210 506 331 520 215 661 331 682 215 519 331 534 220 682 331 704 220 534 331 549 225 704 331 727 225 549 331 566 230 727 331 753 230 566 331 584 235 753 331 780 235 584 331 604 240 780 331 809 240 604 331 625 WBL-21-056 E10 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 209 of 255 Appendix A (Page 10 of 29) 7.0 FIGURES AND TABLES (continued)

Table 2.1-1:

Watts Bar Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate up to 100F/hr) Applicable for 32 EFPY (with the Flange-Notch & without Margins for Instrumentation Errors) Using 1996 App. G Methodology (w/KIA)

(Data points plotted on Figures 2.1-1 and 2.1-2) 60F/hr Heatup Critical Limit 100F/hr Heatup Critical Limit T

[F]

P

[psig]

T

[F]

P

[psig]

T

[F]

P

[psig]

T

[F]

P

[psig]

245 809 331 841 245 625 331 648 250 841 331 875 250 648 331 672 255 875 331 912 255 672 331 699 260 912 331 951 260 699 331 727 265 951 331 993 265 727 331 758 270 993 331 1038 270 758 331 791 275 1038 331 1087 275 791 331 826 280 1087 331 1139 280 826 331 864 285 1139 331 1195 285 864 331 905 290 1195 335 1256 290 905 335 949 295 1256 340 1321 295 949 340 996 300 1321 345 1390 300 996 345 1047 305 1390 350 1465 305 1047 350 1102 310 1465 355 1540 310 1102 355 1160 315 1540 360 1606 315 1160 360 1223 320 1606 365 1677 320 1223 365 1291 325 1677 370 1753 325 1291 370 1363 330 1753 375 1835 330 1363 375 1441 335 1835 380 1922 335 1441 380 1525 WBL-21-056 E11 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 210 of 255 Appendix A (Page 11 of 29) 7.0 FIGURES AND TABLES (continued)

Table 2.1-1:

Watts Bar Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate up to 100F/hr) Applicable for 32 EFPY (with the Flange-Notch & without Margins for Instrumentation Errors) Using 1996 App. G Methodology (w/KIA)

(Data points plotted on Figures 2.1-1 and 2.1-2) 60F/hr Heatup Critical Limit 100F/hr Heatup Critical Limit T

[F]

P

[psig]

T

[F]

P

[psig]

T

[F]

P

[psig]

T

[F]

P

[psig]

340 1922 385 2017 340 1525 385 1614 345 2017 390 2118 345 1614 390 1710 350 2118 395 2227 350 1710 395 1814 355 2227 400 2343 355 1814 400 1925 360 2343 405 2469 360 1925 405 2044 365 2469 365 2044 410 2171 370 2171 415 2308 375 2308 420 2455 380 2455 Leak Test Limit Temp. [F]

310 331 Pressure [psig]

2000 2485 WBL-21-056 E12 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 211 of 255 Appendix A (Page 12 of 29) 7.0 FIGURES AND TABLES (continued)

MATERIAL PROPERTY BASIS LIMITING MATERIAL:

INTERMEDIATE SHELL FORGING 05 LIMITING ART AT 32 EFPY:

1/4-T, 205.74 F 3/4-T, 171.15 F Figure 2.1-3:

Watts Bar Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100F/hr) Applicable for 32 EFPY (with the Flange-Notch &

without Margins for Instrumentation Errors) Using 1996 App. G Methodology (w/KIA)

(Plotted Data (Ref. 1) provided on Table 2.1-2) 0 250 500 750 1000 1250 1500 1750 2000 2250 2500 0

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

Calculated Pressure (PSIG)

Operlim Version:5.2 Run:28921 Operlim.xls Version: 5.2 Unacceptable Operation Acceptable Operation Cooldown Rates

(°F/Hr) steady-state

-20

-40

-60

-100 Boltup Temperature 60 Deg. F WBL-21-056 E13 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 212 of 255 Appendix A (Page 13 of 29) 7.0 FIGURES AND TABLES (continued)

Table 2.1-2:

32 EFPY Cooldown Curve Data Points Using 1996 App. G Methodology (w/KIA, w/Flange Notch & w/o Uncertainties for Instrumentation Errors)

(Data points plotted on Figure 2.1-3)

Steady State 20F/hr 40F/hr 60F/hr 100F/hr T

[F]

P

[psig]

T

[F]

P

[psig]

T

[F]

P

[psig]

T

[F]

P

[psig]

T

[F]

P

[psig]

60 0

60 0

60 0

60 0

60 0

60 513 60 465 60 417 60 367 60 264 65 515 65 467 65 419 65 369 65 266 70 517 70 469 70 421 70 371 70 268 75 519 75 472 75 423 75 373 75 270 80 522 80 474 80 426 80 376 80 272 85 525 85 477 85 428 85 378 85 275 90 528 90 480 90 431 90 381 90 278 95 531 95 483 95 434 95 385 95 281 100 534 100 486 100 438 100 388 100 285 105 538 105 490 105 442 105 392 105 289 110 542 110 494 110 446 110 396 110 294 115 546 115 499 115 450 115 401 115 299 120 550 120 503 120 455 120 406 120 304 125 555 125 508 125 460 125 411 125 310 130 561 130 514 130 466 130 417 130 316 135 566 135 519 135 472 135 423 135 323 140 572 140 526 140 478 140 430 140 331 145 579 145 533 145 485 145 438 145 339 150 586 150 540 150 493 150 446 150 349 WBL-21-056 E14 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 213 of 255 Appendix A (Page 14 of 29) 7.0 FIGURES AND TABLES (continued)

Table 2.1-2:

32 EFPY Cooldown Curve Data Points Using 1996 App. G Methodology (w/KIA, w/Flange Notch & w/o Uncertainties for Instrumentation Errors)

(Data points plotted on Figure 2.1-3)

Steady State 20F/hr 40F/hr 60F/hr 100F/hr T

[F]

P

[psig]

T

[F]

P

[psig]

T

[F]

P

[psig]

T

[F]

P

[psig]

T

[F]

P

[psig]

155 593 155 548 155 501 155 454 155 358 160 602 160 556 160 510 160 464 160 369 165 610 165 565 165 520 165 474 165 381 170 620 170 575 170 530 170 485 170 393 175 630 175 586 175 542 175 497 175 407 180 641 180 597 180 554 180 510 180 422 185 652 185 610 185 567 185 524 185 437 190 665 190 623 190 581 190 539 190 455 195 678 195 637 195 596 195 555 195 473 200 693 200 653 200 613 200 573 200 494 205 708 205 669 205 630 205 592 205 516 210 725 210 687 210 649 210 612 210 539 215 743 215 706 215 670 215 634 215 565 220 763 220 727 220 692 220 658 220 592 225 783 225 749 225 716 225 683 225 622 230 806 230 773 230 741 230 711 230 654 235 830 235 799 235 769 235 741 235 689 240 856 240 827 240 799 240 773 240 726 245 884 245 857 245 831 245 807 245 767 250 913 250 889 250 865 250 844 250 810 WBL-21-056 E15 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 214 of 255 Appendix A (Page 15 of 29) 7.0 FIGURES AND TABLES (continued)

Table 2.1-2:

32 EFPY Cooldown Curve Data Points Using 1996 App. G Methodology (w/KIA, w/Flange Notch & w/o Uncertainties for Instrumentation Errors)

(Data points plotted on Figure 2.1-3)

Steady State 20F/hr 40F/hr 60F/hr 100F/hr T

[F]

P

[psig]

T

[F]

P

[psig]

T

[F]

P

[psig]

T

[F]

P

[psig]

T

[F]

P

[psig]

255 946 255 923 255 903 255 884 255 857 260 980 260 960 260 943 260 928 260 908 265 1017 265 1000 265 986 265 974 265 962 270 1057 270 1043 270 1032 270 1024 270 1021 275 1100 275 1090 275 1082 275 1078 275 1078 280 1147 280 1139 280 1136 280 1136 280 1136 285 1196 285 1193 285 1193 285 1193 285 1193 290 1250 290 1250 290 1250 290 1250 290 1250 295 1307 295 1307 295 1307 295 1307 295 1307 300 1369 300 1369 300 1369 300 1369 300 1369 305 1435 305 1435 305 1435 305 1435 305 1435 310 1507 310 1507 310 1507 310 1507 310 1507 315 1583 315 1583 315 1583 315 1583 315 1583 320 1666 320 1666 320 1666 320 1666 320 1666 325 1755 325 1755 325 1755 325 1755 325 1755 330 1850 330 1850 330 1850 330 1850 330 1850 335 1953 335 1953 335 1953 335 1953 335 1953 340 2063 340 2063 340 2063 340 2063 340 2063 345 2181 345 2181 345 2181 345 2181 345 2181 350 2309 350 2309 350 2309 350 2309 350 2309 WBL-21-056 E16 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 215 of 255 Appendix A (Page 16 of 29) 7.0 FIGURES AND TABLES (continued)

Table 2.1-2:

32 EFPY Cooldown Curve Data Points Using 1996 App. G Methodology (w/KIA, w/Flange Notch & w/o Uncertainties for Instrumentation Errors)

(Data points plotted on Figure 2.1-3)

Steady State 20F/hr 40F/hr 60F/hr 100F/hr T

[F]

P

[psig]

T

[F]

P

[psig]

T

[F]

P

[psig]

T

[F]

P

[psig]

T

[F]

P

[psig]

355 2446 355 2446 355 2446 355 2446 355 2446 WBL-21-056 E17 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 216 of 255 Appendix A (Page 17 of 29) 7.0 FIGURES AND TABLES (continued)

Table 2.1-3 Pressure Differentials Number of Pumps DeltaP (psi) 0 5.2 1

31.0 2

38.0 3

52.0 4

74.0 The COMS analysis considers the following RCP operating restrictions:

RCS Temperature < 105 F, maximum of 2 RCPs in operation RCS Temperature > 105 F, maximum of 4 RCPs in operation WBL-21-056 E18 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 217 of 255 Appendix A (Page 18 of 29) 7.0 FIGURES AND TABLES (continued)

Table 3.1-1 Watts Bar Unit 1 Maximum Allowable COMS PORV Setpoints Indicated Cold Leg Temperature F Maximum Allowable PORV Setpoint (psig) 60 462 105 462 125 481 200 566 225 604 250 664 300 688 350 688 WBL-21-056 E19 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 218 of 255 Appendix A (Page 19 of 29) 7.0 FIGURES AND TABLES (continued)

Table 4-1 Surveillance Capsule Removal Schedule Capsule Capsule Location Lead Factor(a)

Withdrawal EFPY(b)

Fluence (n/cm2)(a)

U 56 5.00 1.20 4.47 x 1018 (c)

W 124 5.05 3.88 1.08 x 1019 (c)

X 236 5.03 6.63 1.71 x 1019 (c)

Z 304 5.06 9.37 2.40 x 1019 (c)

V 58.5 4.31 (d) 6.02 x 1019 Y

238.5 4.31 (e)

Standby (Notes):

(a)

Updated from Capsule Z dosimetry analysis.

(b)

Effective Full Power Years (EFPY) from plant startup.

(c)

Plant specific evaluation.

(d)

Capsule V will be removed during the last scheduled outage before estimated capsule exposure to a neutron fluence equal to two times the peak RPV neutron fluence at 60 years of operation (i.e. 54 EFPY).

(e)

Capsule Y shall remain inserted in the reactor vessel on standby until needed to fulfill future 10 CFR 50, Appendix H or license renewal requirements.

WBL-21-056 E20 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 219 of 255 Appendix A (Page 20 of 29) 7.0 FIGURES AND TABLES (continued)

TABLE 5-1 Comparison of the Watts Bar Unit 1 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions Material Capsule Fluence(d)

(x 1019 n/cm2, E > 1.0 MeV) 30 ft-lb Transition Temperature Shift Upper Shelf Energy Decrease Predicted (F)(a)

Measured (F)(b)

Predicted

(%)(a)

Measured

(%)( c)

Intermediate Shell Forging 05 (Tangential)

U 0.447 95.4 98.3 21 19 W

1.08 125.6 111.4 26 26 X

1.71 141.2 94.7 29 20 Z

2.40 152.0 144.5 31 23 Intermediate Shell Forging 05 (Axial)

U 0.447 95.4 28.7 21 W

1.08 125.6 79.0 26 3.2 X

1.71 141.2 115.9 29 Z

2.40 152.0 104.9 31 0

Surveillance Program Weld Metal U

0.447 31.8 0.0(e) 16 W

1.08 41.9 30.5 19 15 X

1.71 47.1 25.8 22 Z

2.40 50.7 13.9 23 Heat Affected Zone Material U

0.447 50.9 11 W

1.08 48.8 13 X

1.71 74.5 7.9 Z

2.40 67.7 11 WBL-21-056 E21 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 220 of 255 Appendix A (Page 21 of 29) 7.0 FIGURES AND TABLES (continued)

TABLE 5-1 Comparison of the Watts Bar Unit 1 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions (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 5.0.2.

(c)

Values are based on the definition of upper shelf energy given in ASTM E185-82.

(d)

The fluence values presented here are the calculated values.

(e)

Due to the scatter in the Capsule U Weld Charpy test results, a true Hyperbolic Tangent Curve fit resulted in T30 values of -6.4F when compared to unirradiated Charpy test data. A conservative value of 0F was used in RTNDT calculations.

WBL-21-056 E22 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 221 of 255 Appendix A (Page 22 of 29) 7.0 FIGURES AND TABLES (continued)

TABLE 5-2 Calculation of Chemistry Factors using Watts Bar Unit 1 Surveillance Capsule Data Material Capsule Capsule f(a)

FF(b)

RTNDT(c)

FF*RTNDT FF2 Inter. Shell Forging 05 (Tangential)

U 0.447 0.776 98.3 76.28 0.602 W

1.08 1.022 111.4 113.80 1.044 X

1.71 1.148 94.7 108.68 1.317 Z

2.40 1.236 144.5 178.60 1.528 Inter. Shell Forging 05 (Axial)

U 0.447 0.776 28.7 22.27 0.602 W

1.08 1.022 79.0 80.70 1.044 X

1.71 1.148 115.9 133.01 1.317 Z

2.40 1.236 104.9 129.65 1.528 SUM:

842.99 8.981 CF05 = (FF

• RTNDT) (FF2) = (842.99) (8.981) = 93.9F WBL-21-056 E23 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 222 of 255 Appendix A (Page 23 of 29) 7.0 FIGURES AND TABLES (continued)

TABLE 5-2 (Contd)

Calculation of Chemistry Factors using Watts Bar Unit 1 Surveillance Capsule Data Material Capsule Capsule f(a)

FF(b)

RTNDT(c)

FF*RTNDT FF2 Surveillance Weld (d,e,f)

WB1 - U 0.447 0.776 0.0 (0.0(g))

0.00 0.602 WB1 - W 1.08 1.022 40.26 (30.5) 41.13 1.044 WB1 - X 1.71 1.148 34.06 (25.8) 39.08 1.317 WB1 - Z 2.40 1.236 18.35 (13.9) 22.68 1.528 Cat.1 - Z 0.2993 0.670 0.0(h) (1.91) 0.00 0.448 Cat.1 - Y 1.318 1.077 8.52 (17.79) 9.18 1.160 Cat.1 - V 2.334 1.229 15.4 (26.5) 18.93 1.510 MG2 - V 0.323 0.689 32.51 (38.51) 22.41 0.475 MG2 - X 1.47 1.11 29.93 (35.93) 33.13 1.225 MG2 - U 2.04 1.19 17.81 (23.81) 21.27 1.426 MG2 - W 3.07 1.30 37.76 (43.76) 48.94 1.680 SUM:

256.75 12.415 CF Surv. Weld = (FF

• RTNDT) ( FF2) = (256.75) (12.415) = 20.7F WBL-21-056 E24 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 223 of 255 Appendix A (Page 24 of 29) 7.0 FIGURES AND TABLES (continued)

TABLE 5-2 (Contd)

Calculation of Chemistry Factors using Watts Bar Unit 1 Surveillance Capsule Data (Notes):

(a) f = fluence. For Watts Bar Unit 1, see Table 2-2; for Catawba Unit 1, see WCAP-15117, and for McGuire Unit 2, see WCAP-15334. Fluence units are [E+19 n/cm2, E > 1.0 MeV].

(b)

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

(c)

RTNDT values are the measured 30 ft-lb shift values [F]. Watts Bar Unit 1 values are taken from WCAP-16760-NP, Catawba Unit 1 values are taken from WCAP-15117, and McGuire Unit 2 values are taken from WCAP-14799. Units are [F]. The pre-adjusted values are shown in the parentheses; for more details on the adjusted values refer to Notes d and e.

(d)

The Surveillance Weld RTNDT values have been adjusted for temperature difference from Watts Bar Unit 1 as follows:

Minus 7F for Catawba Unit 1 (Note: It is more conservative to adjust for temperature 1st, then apply the ratio.)

Minus 6F for McGuire Unit 2 (e)

The Surveillance Weld RTNDT values have been adjusted for chemistry difference to the best estimate Cu & Ni by a ratio of:

1.32 for Watts Bar Unit 1

0.79 for Catawba Unit 1

1.0 for McGuire Unit 2 (f)

Surveillance data for Catawba Unit 1 and McGuire Unit 2 come from WCAP-15117 and WCAP-14799.

(g)

Original value was -6.4F, but physically a reduction should not occur, therefore a conservative value of zero will be used for Watts Bar Capsule U weld metal.

(h)

After the Temperature adjustment and ratio procedure, the value of the shift comes out less than zero, but physically a reduction should not occur, therefore a conservative value of zero will be used.

WBL-21-056 E25 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 224 of 255 Appendix A (Page 25 of 29) 7.0 FIGURES AND TABLES (continued)

TABLE 5-3 Watts Bar Unit 1 Reactor Vessel Toughness Table (Best Estimate Average, i.e., all data)

Material Description Cu(%)(a)

Ni(%)(a)

Initial RTNDT Closure Head Flange 08 (Heat # 910334 / 710402) 0.13 0.75

-43F Vessel Flange 07 (Heat # 411471)

-40F (d)

Intermediate Shall Forging 05 (Heat # 527536) 0.16 0.80 47F Lower Shell Forging 04 (Heat # 528522) 0.08 0.83 5F Inter. To Lower Shell Girth Weld W05 (b,c) 0.04 0.73

-43F Watts Bar Unit 1 Surveillance Weld Metal (b,c) 0.03 0.75 Watts Bar Unit 2 Surveillance Weld Metal (b,c) 0.02 0.69 Catawba Unit 1 Surveillance Weld Metal (b,c) 0.05 0.73 McGuire Unit 2 Surveillance Weld Metal (b,c) 0.04 0.74 NOTES:

(a)

Based on measured data.

(b)

The surveillance weld was made with the same weld wire and flux as the intermediate to lower shell girth weld (weld wire heat # 895075, flux type Grau L.O. # LW320, lot # P46).

(c)

These values were rounded to two decimal points since the chemistry factor determination uses weight percent chemistry to two decimal points.

(d)

Used in the consideration of flange requirements for heatup/cooldown curves. Per methodology given in WCAP-14040, the minimum boltup temperature is 60F.

WBL-21-056 E26 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 225 of 255 Appendix A (Page 26 of 29) 7.0 FIGURES AND TABLES (continued)

TABLE 5-4 Watts Bar Unit 1 Reactor Vessel Surface Fluence Values at 32 EFPY(a)

Azimuth 0

15 30 45 9.49 x 1018 1.41 x 1019 1.42 x 1019 1.75 x 1019 NOTES:

(a)

These values are based on the new evaluation performed from the Capsule Z evaluation (Ref. 8). Units are n/cm2 (E>1.0 MeV).

TABLE 5-5 Summary of ARTs for the Watts Bar Unit 1 Reactor Vessel Beltline Materials at the 1/4-T and 3/4-T Locations for 32 EFPY 32 EFPY(a)

Component 1/4-T (F) 3/4-T (F)

Intermediate Shell Forging 05 (Heat # 527536) 205.74(b) 171.15(b)

> Using Surveillance Data 176.23 149.82 Lower Shell Forging 04 (Heat # 528522) 90.72 76.38 Intermediate to Lower Shell Girth Weld (Heat # 895075) 66.53 36.15

> Using Surveillance Data

-1.01

-12.66 NOTES:

(a)

Fluence used for the PT Limit curves were based on the previous fluence analysis performed from the Capsule Z evaluation[8] (See Table 5-4 above). Units are n/cm2 (E>1.0 MeV).

(b)

Used to generate the heatup/cooldown curves.

WBL-21-056 E27 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 226 of 255 Appendix A (Page 27 of 29) 7.0 FIGURES AND TABLES (continued)

TABLE 5-6 Calculation of Adjusted Reference Temperatures at 32 EFPY for the Limiting Watts Bar Unit 1 Reactor Vessel Material (Intermediate Shell Forging 05)

Parameter Values Operating Time 32 EFPY Material Inter. Shell Forging 05 Inter. Shell Forging 05 Location 1/4-T 3/4-T Chemistry Factor (CF), F(a) 123 123 Fluence (f), *1019 n/cm2 (E>1.0MeV)(b) 1.05 0.38 Fluence Factor (FF)(c) 1.01 0.73 RTNDT = CF x FF, F 124.74 90.15 Initial RTNDT(I), F 47 47 Margin (M), F(d) 34 34 ART = I + (CF*FF) + M, F

[Per Regulatory Guide 1.99, Revision 2]

205.74 171.15 NOTES:

(a)

CF is based on the Copper and Nickel from Table 5-3.

(b)

Fluence, f, is based upon fsurf = 1.75 x 1019 n/cm2. The Watts Bar Unit 1 reactor vessel wall thickness is 8.465 inches at the beltline region.

(c)

FF = f(0.28 - 0.10 log f)

(d)

Margin is calculated as M = 2(i2 + 2)0.5. The standard deviation for the initial RTNDT margin term, i, is 0F since the initial RTNDT value is a measured value. The standard deviation for the RTNDT margin term, is 17F for the forging except that need not exceed 0.5 times the mean value of RTNDT.

WBL-21-056 E28 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 227 of 255 Appendix A (Page 28 of 29) 7.0 FIGURES AND TABLES (continued)

TABLE 5-7 RTPTS Calculations for Watts Bar Beltline Region Materials at 32 EFPY Material Fluence, f(a)

FF(b)

CF(c)

RTPTS(d)

Margin(e)

RTNDT(U)(f)

RTPTS(g)

Intermediate Shell Forging 05 (Heat # 527536) 1.75 1.15 123 141.88 34 47 222.88 Using Surv. Cap.

Data 1.75 1.15 93.9 108.32 34(i) 47 189.32 Lower Shell Forging 04 (Heat # 528522) 1.75 1.15 51 58.83 34 5

97.83 Inter. To Lower Shell Girth Weld (Heat # 895075) 1.75 1.15 54 62.29 56

-43 75.29 Using Surv. Cap.

Data 1.75 1.15 20.7 23.88 24(h)

-43 4.76 Notes:

(a)

The fluence, f, was taken from the peak azimuthal location (See Table 5-4); *1019 n/cm2, E>1.0 MeV.

(b)

FF = f(28-0.1*logf); where f is the clad/base metal interface fluence.

(c)

The Position 1 (non-surveillance data) Chemistry Factor is obtained from the best estimate Cu & Ni in Table 5-3.

The Position 2 (Surveillance Data) Chemistry Factor is obtained from Table 5-2. Units are F.

(d)

RTPTS = CF*FF; F (e)

Margin = 2*(2 + 2)1/2; F (f)

Initial RTNDT values are measured values (See Table 5-3); F (g)

RTPTS = RTNDT(U) + RTPTS + Margin (This value was rounded per ASTM E29, using the Rounding Method); F (h) exceeded 1/2RTNDT; thus, the calculated value of 1/2RTNDT was used. Surveillance data is credible. See WCAP-16760 (Ref. 8).

(i)

Surveillance Data deemed not-credible, thus the full must be applied. See WCAP-16760 (Ref. 8).

WBL-21-056 E29 of 30

System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record SDD-N3-68-4001 Rev. 0045 Page 228 of 255 Appendix A (Page 29 of 29) 8.0 SOURCE NOTES 1.

NCO820285003 2.

NCO820285004 WBL-21-056 E30 of 30