WBL-25-032, Unit 2 - Revised Pressure and Temperature Limits Reports (PTLR)
| ML25161A019 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 06/10/2025 |
| From: | Reneau W Tennessee Valley Authority |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| WBL-25-032 | |
| Download: ML25161A019 (1) | |
Text
Post Office Box 2000, Spring City, Tennessee 37381 WBL-25-032 June 10, 2025 10 CFR 50.36 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Unit 1 and Unit 2 Facility Operating License Nos. NPF-90 and NPF-96 NRC Docket Nos. 50-390 and 50-391
Subject:
Watts Bar Nuclear Plant Unit 1 and Unit 2 +/- Revised Pressure and Temperature Limits Reports (PTLR)
The purpose of this letter is to provide the enclosed copies of the Watts Bar Unit 1 Pressure and Temperature Limits Report (PTLR), Revision 14, effective May 29, 2025 (Enclosure 1), and the Unit 2 PTLR, Revision 8, effective May 29, 2025 (Enclosure 2), in accordance with Unit 1 and Unit 2 Technical Specifications Section 5.9.6.c.
There are no new regulatory commitments in this letter. Should you have questions regarding this submittal, please contact Jonathan Johnson, Site Compliance Manager, at jtjohnson0@tva.gov.
Respectfully, William C. Reneau Site Vice President Watts Bar Nuclear Plant 1\\14 TENNESSEE VALLEY AUTHORITY
- Reneau, William Christopher Digitally signed by Reneau, William Christopher Date: 2025.06.1 0 05:57:26
-04'00'
U.S. Nuclear Regulatory Commission WBL-25-032 Page 2 June 10, 2025 Enclosures
- 1. Watts Bar Nuclear Plant, Unit 1 Pressure and Temperature Limits Report (PTLR),
Revision 14.
- 2. Watts Bar Nuclear Plant, Unit 2 Pressure and Temperature Limits Report (PTLR),
Revision 8.
cc: (w/ enclosures)
NRC Regional Administrator +/- Region II NRC Senior Resident Inspector - Watts Bar Nuclear Plant NRC Project Manager - Watts Bar Nuclear Plant
WBL-25-032 E1-1 of 1 Watts Bar Nuclear Plant, Unit 1 Pressure and Temperature Limits Report (PTLR),
Revision 14
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 1 of 35)
SDD-Nl-68-4001 Rev.0050 Page 21 O of 271 Watts Bar Unit 1 - RCS Pressure and Temperature Limits Report (PTLR) - Revision 14 APPENDIX "A" TO RCS SYSTEM DESCRIPTION N3-68-4001 WATTS BAR UNIT 1 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
REVISION 14 Prepared by:
J. F. Fitzsimmons Checked by C. J. Zabo Approved by:
C. S. Kerlin
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 2 of 35)
SDD-N3-68-4001 Rev.0050 Page 211 of 271 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 LCO 3.4.12 RCS Pressure and Temperature (PIT) Limits 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 NRG-approved methodologies (Ref. 4, 18, and 19) that are specified in Technical Specification 5.9.6.
2.1 RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3) 2.1.1 The RCS temperature rate-of-change limits are (Ref. 1 ):
A A maximum heatup Rate 100°F per hour.
B.
A maximum cooldown Rate 100°F per hour.
C.
A maximum temperature change of 10°F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
2.1.2 RCS PIT Limits for Heatup, Cooldown, lnservice Hydrostatic and Leak Testing, and Criticality The RCS PIT 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 (0°F 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 +/- 9.2°F (Ref. 1 ).
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 3 of 35)
SDD-N3-68-4001 Rev.0050 Page 212 of 271 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 NRG-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 WCAP-16760-NP (Ref. 8) were 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. The most recent analysis, WCAP-18769-NP (Ref. 14) was used to update 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, RT NDT, which is determined in accordance with ASTM E23. The empirical relationship between RT NDT 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 measured surveillance material 30 ft-lb transition temperature shifts and upper shelf energy decreases.
Table 5-2 provides the Summary of the Watts Bar Unit 1 Reactor Vessel Beltline and Extended Beltline Material Position 1.1 Chemistry Factors.
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 4 of 35) 5.0 SUPPLEMENTAL DATA TABLES (continued)
SDD-N3-68-4001 Rev.0050 Page 213 of 271 Table 5-3 shows the calculation of the Watts Bar Unit 1 Intermediate Shell Forging 05 and Heat #895075 Position 2.1 Chemistry Factors Using Surveillance Capsule Data.
Table 5-4 contains a Summary of the Best Estimate Cu and Ni Weight Percent and Initial RTNDT Values for the Watts Bar Unit 1 Reactor Vessel Beltline and Extended Beltline Materials.
Table 5-5 provides fluence values for the Watts Bar Unit 1 Reactor Vessel Beltline and Extended Beltline Materials at 32 EFPY.
Table 5-6 provides a summary of the limiting Adjusted Reference Temperature values for the Watts Bar Unit 1 Reactor Vessel Beltline and Extended Beltline Materials at the 1/4T and 3/4T locations for 32 EFPY.
Table 5-7 shows the Adjusted Reference Temperature Evaluation for the Watts Bar Unit 1 Reactor Vessel Beltline and Extended Beltline Materials through 32 EFPY at the 1/4T Location.
Table 5-8 shows the Adjusted Reference Temperature Evaluation for the Watts Bar Unit 1 Reactor Vessel Beltline and Extended Beltline Materials through 32 EFPY at the 3/4T Location.
Table 5-9 provides RT PTS values for Watts Bar Unit 1 Beltline and Extended Beltline Materials at 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 3, "Watts Bar Unit 1 Reactor Vessel Radiation Surveillance Program," August 1995.
- 4.
WCAP-14040-A, Revision 4, "Methodology Used To Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 1998.
- 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.
System REACTOR COOLANT SYSTEM Description Document
6.0 REFERENCES
(continued)
Unit 1 / Unit 2 QA Record Appendix A (Page 5 of 35)
SDD-Nl-68-4001 Rev.0050 Page 214 of 271
- 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.
WAT-COM-TR-M-000001, Rev. 0, "Watts Bar Unit 1 COMS Design Basis Update to Include WCAP-14040 Revision 4," Dec. 2022.
- 10.
NE SSD 1-T-68-1 B, COMS Setpoint and Scaling Document for PORV 1-PCV-68-340A Instrument Loop Components.
- 11.
NE SSD 1-T-68-43B, COMS Setpoint and Scaling Document for PORV 1-PCV-68-334 Instrument Loop Components.
- 12.
Deleted.
13 MRP-326, Revision 1, "Coordinated PWR Reactor Vessel Surveillance Program (CRVSP)" June 2021.
14 WCAP-18769-NP, Revision 1, "Watts Bar Units 1 & 2 Reactor Vessel Integrity Evaluations for the 2496 TPBAR Implementation Project," February 2023.
15 NRC Regulatory Issue Summary (RIS) 2014-11, "Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components," U.S. Nuclear Regulatory Commission, October 2014. [Agencywide Documents Access and Management System (ADAMS) Accession Number ML14149A165]
16 WCAP-17669-N P, Revision 1, "Catawba Unit 1 Measurement Uncertainty Recapture (MUR) Power Uprate: Reactor Vessel Integrity and Neutron Fluence Evaluations,"
October 2015.
17 WCAP-17455-NP, Revision 0, "McGuire Units 1 and 2 Measurement Uncertainty Recapture (MUR) Power Uprate: Reactor Vessel Integrity and Neutron Fluence Evaluations," February 2012 18 WCAP-18124-NP-A, Revision 0, "Fluence Determination with RAPTOR-M3G and FERRET," July 2018 19 WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Revision 0, "Fluence Determination with RAPTOR-M3G and FERRET - Supplement for Extended Beltline Materials," May 2022.
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 6 of 35) 7.0 FIGURES AND TABLES SDD-NJ-68-4001 Rev.0050 Page 215 of 271 MATERIAL PROPERTY BASIS LIMITING MATERIAL:
LIMITING ART AT 32 EFPY:
INTERMEDIATE SHELL FORGING 05 1/4-T, 205.74 °F 2500 2250 2000 1750 6'
~ 1500 f
- s
"' "' f 1250
- a.
"C s Ill
]
1000 iii 0
750 500 250 0
3/4-T, 171.15 °F IOperlim Version:5.2 Run:28921 Operlim.xls Version: 5.2 I Leak Test Limit I
~
Unacceptable I Operation I I
I Heatup Rate: I 60 Deg. F/Hr V I
/ /
~..,..,
Temperature 60 Deg. F I
I j
Acceptable ~
I I Operation I
I Critical Limit 60 Deg. F/Hr Criticality Limit based on inservice hydrostatic test temperature (331 F) for the service period up to 32 EFPY 0
50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 2.1-1 :
Watts Bar Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 60°F/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)
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 7 of 35) 7.0 FIGURES AND TABLES (continued)
SDD-NJ-68-4001 Rev.0050 Page 216 of 271 MATERIAL PROPERTY BASIS LIMITING MATERIAL:
LIMITING ART AT 32 EFPY:
INTERMEDIATE SHELL FORGING 05 1/4-T, 205.74 °F 3/4-T, 171.15 °F 2500 IOperlim Version:5.2 Run:28921 Operlim.xls Version: 5.21 I
I 2250 2000 1750 6'
~ 1500 Leak Test Limit I
/
I I
~
Unacceptable I j
Acceptable ~
Operation I
I Operation I I I!!
- I Ill Ill I!! 1250
- a.
"C s Ill 3 1000 u
iii 0
750 500 I I I Critical Limit I
/
100 Deg. F/Hr J
I I
Heatup Rate 100 Deg. F/Hr V
Criticality Limit based on
/
inservice hydrostatic test temperature (331 F) for the service period up to 32 EFPY 250 Boltup I Temperature 60 Deg. F 0
0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 2.1-2:
Watts Bar Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 100°F/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)
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 8 of 35)
SDD-NJ-68-4001 Rev.0050 Page 217 of 271 7.0 FIGURES AND TABLES (continued)
Table 2.1-1 :
60°F/hr Heatup T
p
[OF]
[psig]
60 0
60 465 65 465 70 465 75 465 80 465 85 465 90 465 95 465 100 465 105 465 110 466 115 467 120 470 125 473 130 477 135 481 140 487 145 492 Watts Bar Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate up to 100°F/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)
Critical Limit 100°F/hr Heatup Critical Limit T
p T
p T
p
[OF]
[psig]
[OF]
[psig]
[OF]
[psig]
331 0
60 0
331 0
331 465 60 413 331 413 331 465 65 413 331 413 331 466 70 413 331 414 331 467 75 413 331 414 331 467 80 413 331 416 331 470 85 413 331 416 331 470 90 413 331 418 331 473 95 413 331 419 331 474 100 413 331 421 331 477 105 413 331 422 331 481 110 413 331 425 331 481 115 413 331 427 331 487 120 413 331 429 331 490 125 413 331 433 331 492 130 413 331 435 331 499 135 413 331 440 331 501 140 414 331 441 331 506 145 416 331 447
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 9 of 35)
SDD-NJ-68-4001 Rev.0050 Page 218 of 271 7.0 FIGURES AND TABLES (continued)
Table 2.1-1 :
60°F/hr Heatup T
p
[OF]
[psig]
150 499 155 506 160 514 165 523 170 533 175 543 180 554 185 566 190 579 195 593 200 608 205 625 210 642 215 661 220 682 225 704 230 727 235 753 240 780 Watts Bar Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate up to 100°F/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)
Critical Limit 100°F/hr Heatup Critical Limit T
p T
p T
p
[OF]
[psig]
[OF]
[psig]
[OF]
[psig]
331 514 150 418 331 449 331 516 155 421 331 455 331 523 160 425 331 459 331 533 165 429 331 463 331 543 170 435 331 471 331 554 175 441 331 473 331 566 180 447 331 483 331 579 185 455 331 486 331 593 190 463 331 494 331 608 195 473 331 502 331 625 200 483 331 506 331 642 205 494 331 519 331 661 210 506 331 520 331 682 215 519 331 534 331 704 220 534 331 549 331 727 225 549 331 566 331 753 230 566 331 584 331 780 235 584 331 604 331 809 240 604 331 625
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 1 O of 35)
SDD-NJ-68-4001 Rev.0050 Page 219 of 271 7.0 FIGURES AND TABLES (continued)
Table 2.1-1 :
60°F/hr Heatup T
p
[OF]
[psig]
245 809 250 841 255 875 260 912 265 951 270 993 275 1038 280 1087 285 1139 290 1195 295 1256 300 1321 305 1390 310 1465 315 1540 320 1606 325 1677 330 1753 335 1835 Watts Bar Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate up to 100°F/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)
Critical Limit 100°F/hr Heatup Critical Limit T
p T
p T
p
[OF]
[psig]
[OF]
[psig]
[OF]
[psig]
331 841 245 625 331 648 331 875 250 648 331 672 331 912 255 672 331 699 331 951 260 699 331 727 331 993 265 727 331 758 331 1038 270 758 331 791 331 1087 275 791 331 826 331 1139 280 826 331 864 331 1195 285 864 331 905 335 1256 290 905 335 949 340 1321 295 949 340 996 345 1390 300 996 345 1047 350 1465 305 1047 350 1102 355 1540 310 1102 355 1160 360 1606 315 1160 360 1223 365 1677 320 1223 365 1291 370 1753 325 1291 370 1363 375 1835 330 1363 375 1441 380 1922 335 1441 380 1525
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 11 of 35)
SDD-NJ-68-4001 Rev.0050 Page 220 of 271 7.0 FIGURES AND TABLES (continued)
Table 2.1-1 :
60°F/hr Heatup T
p
[OF]
[psig]
340 1922 345 2017 350 2118 355 2227 360 2343 365 2469 Leak Test Limit Watts Bar Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate up to 100°F/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)
Critical Limit 100°F/hr Heatup Critical Limit T
p T
p T
p
[OF]
[psig]
[OF]
[psig]
[OF]
[psig]
385 2017 340 1525 385 1614 390 2118 345 1614 390 1710 395 2227 350 1710 395 1814 400 2343 355 1814 400 1925 405 2469 360 1925 405 2044 365 2044 410 2171 370 2171 415 2308 375 2308 420 2455 380 2455 Temp. [°F]
310 331 Pressure [psig]
2000 2485
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 12 of 35)
SDD-NJ-68-4001 Rev.0050 Page 221 of 271 7.0 FIGURES AND TABLES (continued)
MATERIAL PROPERTY BASIS LIMITING MATERIAL:
LIMITING ART AT 32 EFPY:
INTERMEDIATE SHELL FORGING 05 1/4-T, 205.74 °F 2500 2250 2000 1750 6'
~ 1500 I!!
- I
- g I!! 1250 CL,,
i
~ 1000 "ii 0
750 500 250 0
3/4-T, 171.15 °F IOperlim Version:5.2 Run:28921 Operlim.xls Version: 5.2 I I
I
~
Unacceptable I Acceptable ~
Operation I
I Operation I
I I I
,I Cooldown Rates C"F/Hr) steady-state
-20
~
-40
-60
-100
~ *~
Boltup I Temperature 60 Deg. F 0
50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 2.1-3:
Watts Bar Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/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)
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 13 of 35)
SDD-N3-68-4001 Rev.0050 Page 222 of 271 7.0 FIGURES AND TABLES (continued)
Table 2.1-2:
Steady State T
p
[OF]
[psig]
60 0
60 513 65 515 70 517 75 519 80 522 85 525 90 528 95 531 100 534 105 538 110 542 115 546 120 550 125 555 130 561 135 566 140 572 145 579 150 586 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) 20°F/hr 40°F/hr 60°F/hr 100°F/hr T
p T
p T
p T
p
[OF]
[psig]
[OF]
[psig]
[OF]
[psig]
[OF]
[psig]
60 0
60 0
60 0
60 0
60 465 60 417 60 367 60 264 65 467 65 419 65 369 65 266 70 469 70 421 70 371 70 268 75 472 75 423 75 373 75 270 80 474 80 426 80 376 80 272 85 477 85 428 85 378 85 275 90 480 90 431 90 381 90 278 95 483 95 434 95 385 95 281 100 486 100 438 100 388 100 285 105 490 105 442 105 392 105 289 110 494 110 446 110 396 110 294 115 499 115 450 115 401 115 299 120 503 120 455 120 406 120 304 125 508 125 460 125 411 125 310 130 514 130 466 130 417 130 316 135 519 135 472 135 423 135 323 140 526 140 478 140 430 140 331 145 533 145 485 145 438 145 339 150 540 150 493 150 446 150 349
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 14 of 35)
SDD-N3-68-4001 Rev.0050 Page 223 of 271 7.0 FIGURES AND TABLES (continued)
Table 2.1-2:
Steady State T
p
[OF]
[psig]
155 593 160 602 165 610 170 620 175 630 180 641 185 652 190 665 195 678 200 693 205 708 210 725 215 743 220 763 225 783 230 806 235 830 240 856 245 884 250 913 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) 20°F/hr 40°F/hr 60°F/hr 100°F/hr T
p T
p T
p T
p
[OF]
[psig]
[OF]
[psig]
[OF]
[psig]
[OF]
[psig]
155 548 155 501 155 454 155 358 160 556 160 510 160 464 160 369 165 565 165 520 165 474 165 381 170 575 170 530 170 485 170 393 175 586 175 542 175 497 175 407 180 597 180 554 180 510 180 422 185 610 185 567 185 524 185 437 190 623 190 581 190 539 190 455 195 637 195 596 195 555 195 473 200 653 200 613 200 573 200 494 205 669 205 630 205 592 205 516 210 687 210 649 210 612 210 539 215 706 215 670 215 634 215 565 220 727 220 692 220 658 220 592 225 749 225 716 225 683 225 622 230 773 230 741 230 711 230 654 235 799 235 769 235 741 235 689 240 827 240 799 240 773 240 726 245 857 245 831 245 807 245 767 250 889 250 865 250 844 250 810
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 15 of 35)
SDD-NJ-68-4001 Rev.0050 Page 224 of 271 7.0 FIGURES AND TABLES (continued)
Table 2.1-2:
Steady State T
p
[OF]
[psig]
255 946 260 980 265 1017 270 1057 275 1100 280 1147 285 1196 290 1250 295 1307 300 1369 305 1435 310 1507 315 1583 320 1666 325 1755 330 1850 335 1953 340 2063 345 2181 350 2309 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) 20°F/hr 40°F/hr 60°F/hr 100°F/hr T
p T
p T
p T
p
[OF]
[psig]
[OF]
[psig]
[OF]
[psig]
[OF]
[psig]
255 923 255 903 255 884 255 857 260 960 260 943 260 928 260 908 265 1000 265 986 265 974 265 962 270 1043 270 1032 270 1024 270 1021 275 1090 275 1082 275 1078 275 1078 280 1139 280 1136 280 1136 280 1136 285 1193 285 1193 285 1193 285 1193 290 1250 290 1250 290 1250 290 1250 295 1307 295 1307 295 1307 295 1307 300 1369 300 1369 300 1369 300 1369 305 1435 305 1435 305 1435 305 1435 310 1507 310 1507 310 1507 310 1507 315 1583 315 1583 315 1583 315 1583 320 1666 320 1666 320 1666 320 1666 325 1755 325 1755 325 1755 325 1755 330 1850 330 1850 330 1850 330 1850 335 1953 335 1953 335 1953 335 1953 340 2063 340 2063 340 2063 340 2063 345 2181 345 2181 345 2181 345 2181 350 2309 350 2309 350 2309 350 2309
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 16 of 35)
SDD-NJ-68-4001 Rev.0050 Page 225 of 271 7.0 FIGURES AND TABLES (continued)
Table 2.1-2:
Steady State T
p
[OF]
[psig]
355 2446 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) 20°F/hr 40°F/hr 60°F/hr 100°F/hr T
p T
p T
p T
p
[OF]
[psig]
[OF]
[psig]
[OF]
[psig]
[OF]
[psig]
355 2446 355 2446 355 2446 355 2446
System REACTOR COOLANT SYSTEM SDD-Nl-68-4001 Description Unit 1 / Unit 2 Rev.0050 Document QA Record Appendix A (Page 17 of 35) 7.0 FIGURES AND TABLES (continued)
Table 2.1-3 Pressure Differentials Page 226 of 271 Number of Pumps Delta0 P (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
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 18 of 35) 7.0 FIGURES AND TABLES (continued)
Table 3.1-1 SDD-Nl-68-4001 Rev.0050 Page 227 of 271 Watts Bar Unit 1 Maximum Allowable COMS PORV Setpoints (Data (Ref. 9) Plotted on Figure 3.1-1)
Indicated RCS Temperature PCV-340A Setpoint (psig)
PCV-334 Setpoint OF (psig) 60 372 377 115 372 377 150 410 415 200 465 470 225 490 495 250 490 545 300 603 705 350 603 705 450 2335 2335
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 19 of 35)
SDD-NJ-68-4001 Rev.0050 Page 228 of 271 7.0 FIGURES AND TABLES (continued) 2500 2000
~ 1500
~
~
~ 1000 500 0
0 Figure 3.1-1 50 PORV #2 Setpoint PORV #l Setpoint 150 200 250
!ndicatedRCS Temp.
300 350 400 450 500 PORV Setpoint vs RCS Temperature (Plotted data (Ref. 9) provided in Table 3.1-1)
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 20 of 35)
SDD-Nl-68-4001 Rev.0050 Page 229 of 271 7.0 FIGURES AND TABLES (continued)
Table 4-1 Surveillance Capsule Removal Schedule Capsule Capsule Lead Removal Time Capsule Fluence Capsule Location Factor<a>
EFPY(b)
(n/cm2)<a>
u 56° 4.87 1.20 4.6 X 1018 w
124° 4.78 3.88 1.08 X 1019 X
236° 4.83 6.62 1.75 X 1019 z
304° 4.76 9.29 2.40 X 1019 V
58.5° 4.11(c) 24.1(d) 5.44 X 1019 y
238.5° 4.oa<c)
(e)
Standby (Notes):
(a)
Capsule lead factors and fluence values are taken from Section 2.2.1 of WCAP-18769-NP, Revision 1 (b)
Effective Full Power Years (EFPY) from plant startup.
(c)
Capsule V lead factor is that projected at the end-of-cycle (EOC) 18, the anticipated withdrawal date. Capsule Y lead factor is calculated at 48 EFPY.
(d)
Projected EFPY at the EOC 18, the anticipated withdrawal date of Capsule V. This removal ensures the capsule exposure remains before two times the peak reactor pressure vessel (RPV) neutron fluence (2.73 x 1019) at 60 years of operation (48 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.
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 21 of 35)
SDD-N3-68-4001 Rev.0050 Page 230 of 271 7.0 FIGURES AND TABLES (continued)
TABLE 5-1 Watts Bar Unit 1 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases Material Capsule Fluence<c>
30 ft-lb Transition Upper Shelf Energy (x 1019 n/cm2, Temperature Shift Decrease E > 1.0 MeV)
Measured Measured (oF)(a)
(o/o)(b)
Intermediate u
0.46 98.3 19 Shell Forging 05 w
1.08 111.4 26 (Tangential)
X 1.75 94.7 20 z
2.40 144.5 23 Intermediate u
0.46 28.7 Shell Forging 05 w
1.08 79.0 3.2 (Axial)
X 1.75 115.9 z
2.40 104.9 0
Surveillance u
0.46 0.0(d)
Program Weld Metal w
1.08 30.5 15 X
1.75 25.8 z
2.40 13.9
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 22 of 35) 7.0 FIGURES AND TABLES (continued)
TABLE 5-1 SDD-N3-68-4001 Rev.0050 Page 231 of 271 Watts Bar Unit 1 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases (Notes}:
(a)
Calculated using measured Charpy data plotted using CVGRAPH, Version 5.0.2.
(b)
Values are based on the definition of upper shelf energy given in ASTM E185-82.
(c)
The fluence values presented here are the "calculated" values.
(d)
Due to the scatter in the Capsule U Weld Charpy test results, a true Hyperbolic Tangent Curve fit resulted in AT 30 values of -6.4°F when compared to unirradiated Charpy test data. A conservative value of 0°F was used in RT Nor calculations.
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 23 of 35) 7.0 FIGURES AND TABLES (continued)
Table 5-2 SDD-N3-68-4001 Rev.0050 Page 232 of 271 Summary of the Watts Bar Unit 1 Reactor Vessel Beltline and Extended Beltline Material Position 1.1 Chemistry Factors Material Description Chemistry Factor (°F)
Reactor Vessel Beltline Materials Intermediate Shell Forging 05 123.0 Lower Shell Forging 04 51.0 Intermediate to Lower Shell Circumferential Weld 54.0 Seam was Reactor Vessel Extended Beltline Materials Upper Shell Forging 06 86.0 Bottom Head Ring 03 37.0 Upper to Intermediate Shell Circumferential Weld 41.0 Seam W06 Lower Shell to Bottom Head Ring Circumferential 41.0 Weld Seam W04 TABLE 5-3 Calculation of the Watts Bar Unit 1 Intermediate Shell Forging 05 and Heat# 895075 Position 2.1 Chemistry Factors using Surveillance Capsule Data Material Capsule Capsule f(a)
FF(b)
L'lRTNDT(c)
FF*L'lRTNDT FF2 u
0.46 0.784 98.3 77.0 0.614 Inter. Shell w
1.08 1.022 111.4 113.8 1.044 Forging 05 (Tangential)
X 1.75 1.154 94.7 109.3 1.331 z
2.40 1.236 144.5 178.6 1.528
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 24 of 35) 7.0 FIGURES AND TABLES (continued) u 0.46 0.784 Inter. Shell w
1.08 1.022 Forging 05 (Axial)
X 1.75 1.154 z
2.40 1.236 SUM:
SDD-N3-68-4001 Rev.0050 Page 233 of 271 28.7 22.5 79.0 80.7 115.9 133.7 104.9 129.7 845.3 CFos = I(FF
- LlRTNor) + I(FF2) = (845.3) + (9.034) = 93.6°F 0.614 1.044 1.331 1.528 9.034
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 25 of 35) 7.0 FIGURES AND TABLES (continued)
TABLE 5-3 (Cont'd)
SDD-N3-68-4001 Rev.0050 Page 234 of 271 Calculation of the Watts Bar Unit 1 Intermediate Shell Forging 05 and Heat# 895075 Position 2.1 Chemistry Factors Using Surveillance Capsule Data Material Capsule Capsule t<a>
FF(b)
LiRTNDT(c)
FF*LiRTNDT FF2 WB1 -U 0.460 0.784 11.88 (0) 9.31 0.614 WB1-W 1.08 1.022 52.14 (30.5) 53.26 1.044 WB1 -X 1.75 1.154 45.94 (25.8) 53.00 1.331 WB1 -Z 2.40 1.236 30.23 (13.9) 37.36 1.528 WB2-U 0.614 0.863 49.92 (32.6) 43.10 0.745 Surveillance Cat.1 -Z 0.286 0.658 10.99 (1.91) 7.23 0.433 Weld (d,e,f)
Cat.1 -Y 1.29 1.071 23.53 (17. 79) 25.20 1.147 Cat.1 -V 2.27 1.222 30.42 (26.5) 37.16 1.493 MG2-V 0.302 0.672 45.51 (38.51) 30.58 0.452 MG2-X 1.38 1.089 42.93 (35.93) 46.77 1.187 MG2-U 1.90 1.176 30.81 (23.81) 36.22 1.382 MG2-W 2.82 1.276 50.76 (43.76) 64.76 1.628 SUM:
443.97 12.983 CF Surv. Weld = I(FF
- llRTNoT) + L( FF2) = (443.97) + (12.893) = 34.2°F
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 26 of 35) 7.0 FIGURES AND TABLES (continued)
TABLE 5-3 (Cont'd)
SDD-N3-68-4001 Rev.0050 Page 235 of 271 Calculation of the Watts Bar Unit 1 Intermediate Shell Forging 05 and Heat# 895075 Position 2.1 Chemistry Factors using Surveillance Capsule Data (Notes}:
(a) f = fluence. For Watts Bar Unit 1, see Table 2-2; for Watts Bar Unit 2, see WCAP-18769 (Ref. 14), for Catawba Unit 1, see WCAP-17669 (Ref. 16), and for McGuire Unit 2, see WCAP-17455 (Ref. 17).
Fluence units are [E+19 n/cm2, E > 1.0 MeV].
(b)
FF= fluence factor= f<0-25 -o.1*109 fl_
(c)
The Surveillance Weld measured ART NOT values have been adjusted for chemistry and irradiation temperature as follows:
Adjusted ART NOT= (ART NOT, Measured+ temp. adjustment) X (CFvesselweld + CFsurv. weld)
The temperature adjustments are based on a time-weighted average temperature of the Watts Bar Unit 1 reactor vessel, which is equal to 557°F over the life of the plant. In addition, the potential of a T avg reduction of 7°F will be taken into consideration; thus, a value of 55O°F will be used for the reactor vessel temperature for the adjustments.
For Watts Bar Unit 1, the CF ratio is 1.32 and temp. adjustment is 9°F (559°F - 55O°F).
For Watts Bar Unit 2, the CF ratio is 1.20 and temp. adjustment is 9°F (559°F - 55O°F).
For Catawba Unit 1, the CF ratio is 0.79 and temp. adjustment is 12°F (562°F - 55O°F).
For McGuire Unit 1, the CF ratio is 1.00 and temp. adjustment is 7°F (557°F - 55O°F).
Pre-adjusted values are shown in parentheses.
System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Document Unit 1 / Unit 2 QA Record Appendix A (Page 27 of 35) 7.0 FIGURES AND TABLES (continued)
TABLE 5-4 Rev.0050 Page 236 of 271 Summary of the Best Estimate Cu and Ni Weight Percent and Initial RT NDT Values for the Watts Bar Unit 1 Reactor Vessel Beltline and Extended Beltline Materials<a>
Material Description Cu wt.%
Niwt. %
Initial RT NDT Closure Head Flange 08 0.13 0.75
-43°F (Heat # 910334 I 710402)
Vessel Flange 07
-4QOF (d)
(Heat # 411471)
Reactor Vessel Beltline Materials Intermediate Shall Forging 05 (Heat# 527536) 0.16 0.80 47°F Lower Shell Forging 04 (Heat# 528522) 0.08 0.83 5°F Inter. To Lower Shell Girth Weld was (Heat# 895075) 0.04 0.73
-43°F Reactor Vessel Extended Beltline Materials Upper Shell Forging 06 (Heat# 411595) 0.12 0.87
-22°F Bottom Head Ring 03 (Heat# 528170) 0.06 0.86
-40°F Upper to Inter. Shell Girth Weld W06 (Heat# 899680) 0.03 0.75 10°F Lower Shell to Bottom Head Ring Girth Weld W04 0.03 0.75 10°F (Heat # 899680)
Reactor Vessel Surveillance Materials Watts Bar Unit 1 Surveillance Weld Metal 0.03 0.75 Watts Bar Unit 2 Surveillance Weld Metal 0.033 0.70 Catawba Unit 1 Surveillance Weld Metal 0.05 0.73 McGuire Unit 2 Surveillance Weld Metal 0.04 0.74 NOTES:
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 28 of 35) 7.0 FIGURES AND TABLES (continued)
SDD-N3-68-4001 Rev.0050 Page 237 of 271 (a)
Values taken from WCAP-18769-NP (Ref. 14), unless otherwise specified. The initial RT Nor values are measured values. The reactor vessel nozzle materials are not considered part of the beltline or extended beltline since the nozzle material fluence values fall below the 1x1017 n/cm2 (E > 1.0 n/cm2) threshold defined by NRC RIS 2014-11 (Ref. 15). Nozzle forging material properties and ART values are detailed in WCAP-18769-NP (Ref. 14).
(b)
Values for the Closure Head Flange and Vessel Flange are taken from WCAP-16761-NP (Ref. 1).
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 29 of 35)
SDD-N3-68-4001 Rev.0050 Page 238 of 271 7.0 FIGURES AND TABLES (continued)
TABLE 5-5 Watts Bar Unit 1 Reactor Vessel Fluence Values at 32 EFPY Material Description Fluence (n/cm2, E > 1.0 MeV)
Clad/Base Metal 1/4T Location 3/4T Location Interface (Inner Surface)
Reactor Vessel Beltline Materials Intermediate Shell 1.71E+19 1.03E+19 3.73E+18 Forging 05 Lower Shell Forging 04 1.75E+19 1.05E+19 3.81E+18 Intermediate to Lower 1.68E+19 1.01 E+19 3.66E+18 Shell Circumferential Weld Seam was Reactor Vessel Extended Beltline Materials Upper Shell Forging 06 5.27E+17 3.17E+17 1.15E+17 Bottom Head Ring 03 1.19E+18 7.16E+17 2.59E+17 Upper to Intermediate 6.25E+17 3.76E+17 1.36E+17 Shell Circumferential Weld Seam W06 Lower Shell to Bottom 1.43E+18 8.61E+17 3.12E+17 Head Ring Circumferential Weld SeamW04 TABLE 5-6 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 Limiting ART (°F)(a)
(Intermediate Shell Forging 05)
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 30 of 35) 7.0 FIGURES AND TABLES (continued) 1/4T 205.Q(b)
NOTES:
SDD-N3-68-4001 Rev.0050 Page 239 of 271 3/4T 17Q.4(b)
(a)
The Limiting ART values were calculated in WCAP-18769-NP (Ref. 14).
(b)
The actual ART values (1/4T:205.74°F, 3/4T:171.15°F) used to generate the heatup and cooldown curves were developed in WCAP-16760-NP (Ref. 8).
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 31 of 35) 7.0 FIGURES AND TABLES (continued)
TABLE 5-7 SDD-N3-68-4001 Rev.0050 Page 240 of 271 Adjusted Reference Temperature Evaluation for the Watts Bar Unit 1 Reactor Vessel Beltline and Extended Beltline Materials through 32 EFPY at the 1/4T Location(a)
Reactor Vessel CF 1/4T f 1/4T RTNDT(U)
Predicted Oi 01',.
M ART Location (OF)
(x101e FF (OF)
L'.\\RTNDT (OF)
(OF)
(OF)
(OF) n/cm2, E >
(OF) 1.0 MeV)
Reactor Vessel Beltline Materials Intermediate 123.0 1.03 1.008 47 LJ 0.0 17.0 34.0 205.0 Shell Forging 05 Using Non-93.6 1.03 1.008 47 94.3 0.0 17.0 34.0 175.3 Credible Surveillance Data Lower Shell 51.0 1.05 1.014 5
51.7 0.0 17.0 34.0 90.7 Forging 04 Intermediate to 54.0 1.01 1.003
-43 54.2 0.0 27.1 54.2 65.3 Lower Shell Circumferential Weld Seam was Using Credible 34.2 1.01 1.003
-43 34.3 0.0 14.0 28.0 19.3 Surveillance Data Reactor Vessel Extended Beltline Materials Upper Shell 86.0 0.0317 0.227
-22 19.5 0.0 9.8 19.5 17.0 Forging 06 Bottom Head 37.0 0.0716 0.353
-40 ~
0.0 6.5 13.1
-13.8 Ring 03
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 32 of 35) 7.0 FIGURES AND TABLES (continued)
Upper to 41.0 0.0376 0.250 10 Intermediate Shell Circumferential weld Seam W06 Lower Shell to 41.0 0.0861 0.387 10 Bottom Head Ring Weld Seam W04 NOTES:
(a)
Values are taken from WCAP-18769-NP (Ref. 14).
TABLE 5-8 SDD-N3-68-4001 Rev.0050 Page 241 of 271 10.3 0.0 5.1 10.3 30.5 15.9 0.0 7.9 15.9 41.8 Adjusted Reference Temperature Evaluation for the Watts Bar Unit 1 Reactor Vessel Beltline and Extended Beltline Materials through 32 EFPY at the 1/4T Location(a)
Reactor Vessel CF 3/4Tf 3/4T RTNDT(U)
Predicted Oi 01',.
M ART Location (OF)
(x101e FF (OF)
L'.\\RTNDT (OF)
(OF)
(OF)
(OF) n/cm2, E>
(OF) 1.0MeV)
Reactor Vessel Beltline Materials Intermediate 123.
0.373 0.727 47 [:]
0.0 17.0 34.0 170.4 Shell Forging 05 0
Using Non-93.6 0.373 0.727 47 68.1 0.0 17.0 34.0 149.1 Credible Surveillance Data Lower Shell 51.0 0.381 0.733 5
37.4 0.0 17.0 34.0 76.4 Forging 04 Intermediate to 54.0 0.366 0.722
-43 39.0 0.0 19.5 39.0 35.0 Lower Shell Circumferential Weld Seam was
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 33 of 35)
SDD-N3-68-4001 Rev.0050 Page 242 of 271 7.0 FIGURES AND TABLES (continued)
Using Credible EJLJELJLJWEJLJ Surveillance Data Reactor Vessel Extended Beltline Materials Upper Shell I BS.O II~ II o.,zo I =:Jc:JTIEJB Forging 06 Bottom Head E][::JEI:J[:JEI:][:][:]
Ring 03 Upper to 41.0 0.0136 0.135 10 5.5 0.0 2.8 5.5 21.0 Intermediate Shell Circumferential Weld Seam W06 Lower Shell to 41.0 0.0312 0.225 10 9.2 0.0 4.6 9.2 28.4 Bottom Head Ring Weld SeamW04 Notes:
(a)
Values are taken from WCAP-18769-NP (Ref. 14)
Table 5-9 RT PTs Calculations for the Watts Bar Unit 1 Reactor Vessel Beltline and Extended Beltline Materials through 32 EFPY(a)
Reactor Vessel CF (°F)
Predicted Oi 0~ (°F)
M RTPTs Location Fluence (OF)
ARTNDT (OF)
(OF)
(OF)
(x101s (OF) n/cm2, E >
1.0 MeV Reactor Vessel Beltline Materials Intermediate 123.0 1.71 1.148 47 141.2 0.0 17.0 34.0 222.2 Shell Forging 05
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix A (Page 34 of 35)
SDD-N3-68-4001 Rev.0050 Page 243 of 271 7.0 FIGURES AND TABLES (continued)
Using Non-93.6 1.71 1.148 47 107.4 0.0 17.0 34.0 Credible Surveillance Data Lower Shell 51.0 1.75 1.154 5
58.8 0.0 17.0 34.0 Forging 04 Intermediate to 54.0 1.68 1.143
-43 61.7 0.0 28.0 56.0 Lower Shell Circumferential Weld Seam was Using Credible 34.2 1.68 1.143
-43 39.1 0.0 14.0 28.0 Surveillance Data Reactor Vessel Extended Beltline Materials Upper Shell 86.0 0.0527 0.301
-22 25.9 0.0 12.9 25.9 Forging 06 Bottom Head 37.0 0.119 0.453
-40 16.7 0.0 8.4 16.7 Ring 03 Upper to 41.0 0.0625 0.330 10 13.5 0.0 6.8 13.5 Intermediate Shell Circumferential Weld Seam W06 Lower Shell to 41.0 0.143 0.492 10 20.2 0.0 10.1 20.2 Bottom Head Ring Weld Seam W04 NOTES (a)
Values are taken from WCAP-18769-NP (Ref. 14).
188.4 97.8 74.7 24.1 29.8
-6.5 37.0 50.4
System REACTOR COOLANT SYSTEM Description Document Unit 1 / Unit 2 QA Record Appendix A (Page 35 of 35) 7.0 FIGURES AND TABLES (continued) 8.0 SOURCE NOTES
- 1.
NCO820285003
- 2.
NCO820285004 SDD-N3-68-4001 Rev.0050 Page 244 of 271
WBL-25-032 E2-1 of 1 Watts Bar Nuclear Plant, Unit 2 Pressure and Temperature Limits Report (PTLR),
Revision 8
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix B (Page 1 of 21)
SDD-Nl-68-4001 Rev.0050 Page 245 of 271 Watts Bar Unit 2 - RCS Pressure and Temperature Limits Report (PTLR) - Revision 8 APPENDIX "B" TO RCS SYSTEM DESCRIPTION N3-68-4001 WATTS BAR UNIT 2 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
REVISION 8 Prepared by:
J. F. Fitzsimmons Checked by:
C. J. Zabo Approved by:
C. S. Kerlin
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix B (Page 2 of 21)
SDD-N3-68-4001 Rev.0050 Page 246 of 271 1.0 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
This PTLR for Watts Bar Unit 2 has been prepared in accordance with the requirements of Technical Specification 5.9.6. Revisions to the PTLR shall be provided to the NRC within 30 days of issuance.
The Technical Specifications affected by this report are listed below:
LCO 3.4.3, RCS Pressure and Temperature (PIT) 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 specified in Technical Specification 5.9.6.
2.1 RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3) 2.1.1 The minimum boltup temperature is 60°F.
2.1.2 The RCS temperature rate-of-change limits are:
A A maximum heatup rate of 100°F per hour.
B.
A maximum cooldown rate of 100°F per hour.
C.
A maximum temperature change of~ 10°F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
2.1.3 RCS PIT Limits for Heatup, Cooldown, lnservice Hydrostatic and Leak Testing, and Criticality The RCS PIT limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticality are specified by Figures 2.1-1 and 2.1-2 (Ref. 1 ).
3.0 COLD OVERPRESSURE MITIGATION SYSTEM (LCO 3.4.12)
The lift setting limits for the pressurizer Power Operated Relief Valves (PORVs) are presented in the subsection that follows. These lift setting limits 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 setting limits are specified by Figure 3.1-1 and Table 3.1-1 (Ref. 2).
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix B (Page 3 of 21)
SDD-Nl-68-4001 Rev.0050 Page 247 of 271 NOTE:
These setpoints include allowance for pressure difference between the pressure transmitter and reactor midplane, and also includes a 71.8 psig pressure channel uncertainty, and a 16.3°F temperature uncertainty.
3.2 Arming Temperature COMS shall be armed when any RCS cold leg temperature is ~ 225°F for Unit 2.
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 results of these examinations shall be used to update Figures 2.1-1, 2.1-2, and 3.1-1.
The pressure vessel steel surveillance program (Ref. 3) is in compliance with Appendix H to 10 CFR 50 (Ref. 4), entitled "Reactor Vessel Material Surveillance Program Requirements."
The material test requirements and the acceptance standard utilize the reference nil-ductility temperature, RT NDT, which is determined in accordance with ASTM E208 (Ref. 5). The empirical relationship between RT NDT 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 (Ref. 6). The surveillance capsule removal schedule meets the requirements of ASTM E185-82 (Ref. 7).
The removal schedule is provided in Table 4.0-1.
To date, one surveillance capsule has been removed from Watts Bar Unit 2, Capsule U. The testing results from Capsule U are documented in WCAP-18518-NP (Ref. 10). Regulatory Guide (RG) 1.99, Revision 2, requires at least two credible data sets for use in embrittlement calculations. Since Capsule U is the first capsule to be withdrawn from Watts Bar Unit 2, this criterion is not satisfied for the surveillance forging. However, there is additional surveillance data available from sister plants for the surveillance weld. The Watts Bar Unit 1, Catawba Unit 1, and McGuire Unit 2 surveillance programs contain weld wire Heat# 895075, which was also used in the fabrication of Watts Bar Unit 2 intermediate to lower shell circumferential Weld Seam W05. Thus, the data from these surveillance programs are applicable to Watts Bar Unit 2, and the RG 1.99, Position 2.1 chemistry factor may be used to make embrittlement projections. Table 5-4 contains the sister-plant weld material surveillance data as well as the data from Watts Bar Unit 2, Capsule U, surveillance weld.
The effect of these results on Figures 2.1-1, 2.1-2, and 3.1-1 were evaluated in WCAP-18532-N P (Ref. 11) and L TR-SCS-20-53 (Ref 12).
5.0 SUPPLEMENTAL DATA TABLES Table 5-1 contains a Summary of the Best Estimate Cu and Ni Weight Percent and Initial RT NDT Values for the Watts Bar Unit 2 Reactor Vessel Beltline and Extended Beltline Materials.
Table 5-2 shows a Summary of the Initial RT NDT Values for the Watts Bar Unit 2 Closure Head and Vessel Flange.
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix B (Page 4 of 21)
SDD-Nl-68-4001 Rev.0050 Page 248 of 271 Table 5-3 provides the Summary of the Watts Bar Unit 2 Reactor Vessel Beltline and Extended Beltline Material Position 1.1 Chemistry Factors.
Table 5-4 provides the Watts Bar Unit 2, Catawba Unit 1, Watts Bar Unit 1, and McGuire Unit 2 Surveillance Weld Data for Heat #895075 Table 5-5 shows the Calculation of the Watts Bar Unit 2 Heat #895075 Position 2.1 Chemistry Factor Using Surveillance Capsule Data.
Table 5-6 provides Fluence Values for the Watts Bar Unit 2 Reactor Vessel Beltline and Extended Beltline Materials.
Table 5-7 shows Adjusted Reference Temperature Evaluation for the Watts Bar Unit 2 Reactor Vessel Beltline and Extended Beltline Materials through 32 EFPY at the 1/4T Location.
Table 5-8 contains Adjusted Reference Temperature Evaluation for the Watts Bar Unit 2 Reactor Vessel Beltline and Extended Beltline Materials through 32 EFPY at the 3/4T Location.
Table 5-9 provides a Summary of the Limiting ART Values Used in the Generation of the Watts Bar Unit 2 Heatup/Cooldown Curves.
Table 5.10 shows RT PTS calculations for the Watts Bar Unit 2 Beltline and Extended Beltline Materials at 32 EFPY.
6.0 REFERENCES
- 1.
WCAP-18191-NP, Revision 1, 'Watts Bar Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and Supplemental Reactor Vessel Integrity Evaluations,"
February 2020.
- 2.
Westinghouse Letter L TR-SCS-17-34, Revision 0, "Watts Bar Unit 2 Cold Overpressure Mitigation System (COMS) Setpoint Analysis due to TPBARs" dated August 14, 2017.
- 3.
WCAP-9455, Revision 4, "Tennessee Valley Authority Watts Bar Unit No. 2 Reactor Vessel Radiation Surveillance Program," August 2019.
- 4.
Code of Federal Regulations, 10 CFR 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements," U.S. Nuclear Regulatory Commission, Federal Register, Volume 60, No. 243, December 19, 1995.
- 5.
ASTM E208, "Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels," American Society for Testing and Materials.
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix B (Page 5 of 21)
SDD-Nl-68-4001 Rev.0050 Page 249 of 271
- 6.
Appendix G to the 1998 through the 2000 Addenda Edition of the ASME Boiler and Pressure Vessel Code,Section XI, Division 1, "Fracture Toughness Criteria for Protection Against Failure."
- 7.
ASTM E185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," E706 (IF), ASTM 1982.
- 8.
WCAP-13830, Revision 1, "Heat Up and Cool Down Limit Curves for Normal Operation for Watts Bar Unit 2," J. M. Chicots, et al, February 1995.
9 NRC Regulatory Issue Summary (RIS) 2014-11, "Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components," U.S. Nuclear Regulatory Commission, October 2014. [Agencywide Documents Access and Management System (ADAMS)
Accession Number ML14149A165]
10 WCAP-18518-NP, Revision 0, "Analysis of Capsule U from the Watts Bar Unit 2 Reactor Vessel Radiation Surveillance Program," March 2020 11 WCAP-18532-NP, Revision 0, 'Watts Bar Unit 2 Reactor Vessel Integrity Evaluations," August 2020 12 Westinghouse Letter L TR-SCS-20-53, Revision 0, "Watts Bar Unit 2 Cold Overpressure Mitigation System (COMS) Evaluation for Capsule U," dated September 28, 2020.
13 WCAP-18769-NP, Revision 1, 'Watts Bar Units 1 & 2 Reactor Vessel Integrity Evaluations for the 2496 TPBAR Implementation Project," February 2023.
14 WCAP-17455-NP, Revision 0, "McGuire Units 1 and 2 Measurement Uncertainty Recapture (MUR) Power Uprate: Reactor Vessel Integrity and Neutron Fluence Evaluations," February 2012 15 CAP-17669-NP, Revision 1, "Catawba Unit 1 Measurement Uncertainty Recapture (MUR) Power Uprate: Reactor Vessel Integrity and Neutron Fluence Evaluations,"
October 2015.
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix B (Page 6 of 21)
SDD-NJ-68-4001 Rev.0050 Page 250 of 271 7.0 FIGURES AND TABLES MATERIAL PROPERTY BASIS LIMITING MATERIAL: Intermediate Shell Forging 05 using Reg. Guide 1.99 Position 1.1 LIMITING ART VALUES AT 32 EFPY:
1/4T, 88°F (Axial Flaw) 3/4T, 71 °F (Axial Flaw) 2500 ~~- -- _- _--:.,_- _- _- _- _- _- _- _- _- _- _- _- _--:.,_- _- _- _- _- _- _- _- _--:.,_- _- _- ~--------~
2250 2000 1750
§'
[
1500 41...
~ 1250 a..
-0 41..
~
- , 1000
~
u 750 500 250 OperlimAnalysis Version:5.4 R un:29007 Operlim.xis m Version: 5.4.1 Leak Test Limit Unacceptable 0 er.tion l
Critic. I Limit 60°F/Hr Critical Limit 100°F/Hr AcceJ>table 0 >eration Criticality Limit based on inseivice hydrostatic test temperature (149'1=} for the service period up to 32 EFPY 0 +-r-......-,.-++-r-r....+-r-r-,~"T""T""......-li-T-r-r-r+-r-r-r-,-+-,-"T""T""..-+-,.......-.--r+...-r-......-1--.-r-r-r-+-r-.........-t 0
50 100 150 200 250 300 350 400 450 500 550 Moder.1tor Temper.1ture (Deg. F)
Figure 2.1-1 Watts Bar Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100°F/hr)
Applicable for 32 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ Kie)
(Plotted data (Ref. 1) provided in Table 2.1-1)
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix B (Page 7 of 21)
TABLE 2.1-1 Watts Bar Unit 2 Heatup Limits SDD-Nl-68-4001 Rev.0050 Page 251 of 271 32 EFPY Heatup Curve Data Points Using 1998 through 2000 Addenda App. G Methodology Data (Ref. 1) plotted on Figure 2.1-1 LEAK HEATUP CRITICALITY HEATUP CRITICALITY TEST RATE LIMITS RATE LIMITS LIMITS (60°F/HR)
(60°F/HR)
(100°F/HR)
(100°F/HR)
T p
T p
T p
T p
T p
(OF)
(psig)
(OF)
(psig)
(OF)
(psig)
(OF)
(psig)
(OF)
(psig) 132 2000 60 0
149 0
60 0
149 0
149 2485 60 621 149 1021 60 621 149 904 65 621 150 1032 65 621 150 909 70 621 155 1076 70 621 155 934 75 621 160 1126 75 621 160 963 80 621 165 1183 80 621 165 996 85 621 170 1246 85 621 170 1035 90 621 175 1317 90 621 175 1079 95 621 180 1395 95 621 180 1129 100 621 185 1483 100 621 185 1185 100 960 190 1580 100 873 190 1248 105 993 195 1687 105 889 195 1318 110 1032 200 1806 110 909 200 1396 115 1076 205 1938 115 934 205 1483 120 1126 210 2083 120 963 210 1580 125 1183 215 2244 125 996 215 1686 130 1246 220 2421 130 1035 220 1805 135 1317 135 1079 225 1936 140 1395 140 1129 230 2080 145 1483 145 1185 235 2240 150 1580 150 1248 240 2417 155 1687 155 1318 160 1806 160 1396 165 1938 165 1483 170 2083 170 1580 175 2244 175 1686 180 2421 180 1805
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix B (Page 8 of 21)
TABLE 2.1-1 Watts Bar Unit 2 Heatup Limits SDD-Nl-68-4001 Rev.0050 Page 252 of 271 32 EFPY Heatup Curve Data Points Using 1998 through 2000 Addenda App. G Methodology Data (Ref. 1) plotted on Figure 2.1-1 LEAK HEATUP CRITICALITY HEATUP CRITICALITY TEST RATE LIMITS RATE LIMITS LIMITS (60°F/HR)
(60°F/HR)
(100°F/HR)
(100°F/HR)
T p
T p
T p
T p
T p
(OF)
(psig)
(OF)
(psig)
(OF)
(psig)
(OF)
(psig)
(OF)
(psig) 185 1936 190 2080 195 2240 200 2417
System REACTOR COOLANT SYSTEM SDD-NJ-68-4001 Description Unit 1 / Unit 2 Rev.0050 Document QA Record Page 253 of 271 Appendix B (Page 9 of 21)
MATERIAL PROPERTY BASIS LIMITING MATERIAL: Intermediate Shell Forging 05 using Reg. Guide 1.99 Position 1.1 LIMITING ART VALUES AT 32 EFPY:
1/4T, 88°F (Axial Flaw) 3/4T, 71 °F (Axial Flaw) 2500 i:=============================:::;---i----.---i joperlimAnalysis Vers ion:5.4 R un:29907 Operlim.xl>m Version: 5.4.1 I 2250 2000 1750 G'
~ 1500 41...
~ 1250 Q.
-0 41
~
- , 1000
~
u 750 500 250 t
Unacceptable 0 >eration Cooldow11 Rates:
steady-state
-20°F/Hr
-40°F/Hr
-60°F/Hr
-100°F/Hr Accer>table 0 >eration 0.................,......................-+-..-.-.-.-+............-.--11-.--r.......... +-.-..-.-.-+-,-..,.........-+-,.-.--,.......+-..-.-......+-.-.--.-.-+-,-,,~
0 50 100 150 200 250 300 350 400 450 500 550 Mod er.ito r Tern per.iture {Deg. F)
Figure 2.1-2 Watts Bar Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60, and 100°F/hr) Applicable for 32 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ Kie)
(Plotted data (Ref. 1) provided in Table 2.1-2)
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix B (Page 10 of 21)
TABLE 2.1-2 Watts Bar Unit 2 Cooldown Limits SDD-Nl-68-4001 Rev.0050 Page 254 of 271 32 EFPY Heatup Curve Data Points Using 1998 through 2000 Addenda App. G Methodology (Data (Ref. 1) plotted on Figure 2.1-2)
Steady State 20°F/HR 40°F/HR 60°F/HR 100°F/HR T
p T
p T
p T
p T
p (OF)
(psig)
(OF)
(psig)
(OF)
(psig)
(OF)
(psig)
(OF)
(psig) 60 0
60 0
60 0
60 0
60 0
60 621 60 621 60 621 60 621 60 621 65 621 65 621 65 621 65 621 65 621 70 621 70 621 70 621 70 621 70 621 75 621 75 621 75 621 75 621 75 621 80 621 80 621 80 621 80 621 80 621 85 621 85 621 85 621 85 621 85 621 90 621 90 621 90 621 90 621 90 621 95 621 95 621 95 621 95 621 95 621 100 621 100 621 100 621 100 621 100 621 100 1080 100 1075 100 1075 100 1075 100 1075 105 1130 105 1130 105 1130 105 1130 105 1130 110 1185 110 1185 110 1185 110 1185 110 1185 115 1247 115 1247 115 1247 115 1247 115 1247 120 1315 120 1315 120 1315 120 1315 120 1315 125 1390 125 1390 125 1390 125 1390 125 1390 130 1473 130 1473 130 1473 130 1473 130 1473 135 1564 135 1564 135 1564 135 1564 135 1564 140 1665 140 1665 140 1665 140 1665 140 1665 145 1777 145 1777 145 1777 145 1777 145 1777 150 1901 150 1901 150 1901 150 1901 150 1901 155 2037 155 2037 155 2037 155 2037 155 2037 160 2188 160 2188 160 2188 160 2188 160 2188 165 2355 165 2355 165 2355 165 2355 165 2355
System Description Document 2500 2000 1500
-~
~
t:
'15...
~
~
"- 1000 500 0
60 110 REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix B (Page 11 of 21)
Setpoint Window SDD-NJ-68-4001 Rev.0050 Page 255 of 271 I
I I
,-----------------J PVC-334 PCV-340A I 160 210 260 310 Indicated RCS Temperature ("F)
Figure 3.1-1 PORV Setpoint vs RCS Temperature (Plotted data (Ref. 2) provided in Table 3.1-1) 360 410
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix B (Page 12 of 21)
TABLE 3.1-1 SDD-Nl-68-4001 Rev.0050 Page 256 of 271 Watts Bar Unit 2 PORV Setpoints vs Temperature (Data (Ref. 2) Plotted on Figure 3.1-1)
Temperature PCV-334 Setpoint PCV-340A Setpoint (OF)
(psig)
(psig) 60 416 409 117 416 409 125 490 483 167 490 483 190 696 590 225 696 590 300 696 590 350 696 590 450 2335 2335
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix B (Page 13 of 21)
TABLE 4.0-1 SDD-Nl-68-4001 Rev.0050 Page 257 of 271 Watts Bar Unit 2 Surveillance Capsule Removal Schedule <a>
Capsule Capsule Capsule Removal Expected Capsule Location Lead Time Fluence Factor<a>
{EFPY)<bJ (n/cm2,E > 1.0 MeV)<a>
u Dual 34° 4.80 2.00 EFPY 0.614 X 1019 (EOC 2) w Single 34° 4.87 6.8 EFPY<c>
1.93 X 1019 X
Dual 34°
~4.8 6.8 EFPYto 1.93 X 1019 to 13.8 EFPY(d) 3.86 X 1019 (d) z Single 34° 4.55 Standby<e>
Standby<e>
V Dual 31.5° 3.94 Standby<e>
Standby<e>
y Dual 31.5° 3.94 Standby<e>
Standby<e>
Notes:
(a) Capsule lead factors and fluence values are taken from Section 2.2.2 of WCAP-18769-NP (Ref. 13).
(b)
Effective full power years (EFPY) from plant startup.
(c) Capsule W should be withdrawn at the outage nearest to but following 6.8 EFPY of operation.
(d) Capsule X should be removed between 10.4 EFPY and 13.8 EFPY if possible. Capsule X must be withdrawn between 6.8 EFPY and 13.8 EFPY in order to satisfy the recommendations of the third capsule for EOL per ASTM E185-82. The removal EFPY of the third capsule should be revisited at a later date, such as after Capsule Wis removed.
(e) Capsules Z, V, and Y should remain in the reactor. If additional metallurgical data is needed, withdrawal and testing of these capsules should be considered. Per ASTM E185-82 and NUREG 1801, Revision 2, it is recommended that the capsules be removed prior to reaching a fluence of two times the peak fluence at EOL. In the event that Capsule W cannot be withdrawn, Capsule Z may be removed instead, during the same outage, to satisfy the ASTM E185-82 requirements for the second withdrawn capsule.
System REACTOR COOLANT SYSTEM SDD-Nl-68-4001 Description Unit 1 / Unit 2 Rev.0050 Document QA Record Page 258 of 271 Appendix B (Page 14 of 21)
Table 5-1 Summary of the Best Estimate Cu and Ni Weight Percent and Initial RT NDT Values for the Watts Bar Unit 2 Reactor Vessel Beltline and Extended Beltline Materials<a>
Material Description Chemical Composition Initial Cu Ni RTNDT wt.%
wt.%
Reactor Vessel Beltline Materials Intermediate Shell Forging 05 0.05 0.78 14°F Lower Shell Forging 04 0.05 0.81 5°F Intermediate to Lower Shell Circumferential 0.04 0.73
-50°F Weld Seam was (Heat #895075)
Reactor Vessel Extended Beltline Materials Upper Shell Forging 06 0.07 0.91
-14°F Upper to Intermediate Shell Circumferential 0.03 0.75 10°F Weld Seam W06 (Heat #899680)
Lower Shell to Bottom Head Ring 0.03 0.75 10°F Circumferential Weld Seam W04 (Heat
- 899680)
Bottom Head Ring 03 0.06 0.86
-40°F Note:
(a) Values taken from WCAP-18769-NP (Ref. 13). The initial RT NDT values are measured values. The reactor vessel nozzle materials are not considered part of the beltline or extended beltline since the nozzle material fluence values fall below the 1 x 1017 n/cm2 (E > 1.0 n/cm2) threshold defined by NRC RIS 2014-11 (Ref. 9).
Table 5-2 Summary of the Initial RT NDT Values for the Watts Bar Unit 2 Closure Head and Vessel Flange Material Identification Initial RT NDT<a>
Closure Head Flange
-40°F Vessel Flange
-22°F Note:
(a) The initial RTNoT values are measured values, taken from WCAP-18191-NP (Ref. 1) and consistent with WCAP-13830, Revision 1 (Ref. 8)
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix B (Page 15 of 21)
SDD-Nl-68-4001 Rev.0050 Page 259 of 271 Table 5-3 Summary of the Watts Bar Unit 2 Reactor Vessel Beltline and Extended Beltline Material Position 1.1 Chemistry Factors Material Description Position 1.1 Chemistry Factor Reactor Vessel Beltline Materials Intermediate Shell Forging 05 31.0°F Lower Shell Forging 04 31.0°F Intermediate to Lower Shell 54.0°F Circumferential Weld Seam W05 Reactor Vessel Extended Beltline Materials Upper Shell Forging 06 44.0°F Upper to Intermediate Shell Circumferential Weld 41.0°F Seam W06 Lower Shell to Bottom Head Ring Circumferential 41.0°F Weld Seam W04 Bottom Head Ring 03 37.0°F Table 5-4 Watts Bar Unit 2, Catawba Unit 1, Watts Bar Unit 1, and McGuire Unit 2 Surveillance Weld Data for Heat #895075<a>
Material Capsule Cu Ni Irradiation Capsule (wt.%)
(wt.%)
Temperature fluence (Of)
(x 1019 n/cm2, E > 1.0 MeV)
Watts Bar u
0.033 0.70 559 0.614 Unit2 Catawba z
0.05 0.73 562 0.286 Unit 1 y
0.05 0.73 562 1.29 V
0.05 0.73 562 2.27 Watts Bar u
0.03 0.75 559 0.46 Unit 1 w
0.03 0.75 559 1.08 X
0.03 0.75 559 1.75 z
0.03 0.75 559 2.40 McGuire V
0.04 0.74 557 0.302 Unit2 X
0.04 0.74 557 1.38 u
0.04 0.74 557 1.90 w
0.04 0.74 557 2.82 Note:
(a) Surveillance data taken from WCAP-18769-NP (Ref. 13).
ARTNDT (Of) 32.6 1.91 17.79 26.5 0.0 30.5 25.8 13.9 38.51 35.93 23.81 43.76
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix B (Page 16 of 21)
SDD-Nl-68-4001 Rev.0050 Page 260 of 271 Table 5-5 Calculation of the Watts Bar Unit 2 Heat# 895075 Position 2.1 Chemistry Factor Using Surveillance Capsule Data Material Capsule Capsule f(aJ ff(b)
ARTNDT(c)
FF* ARTNDT FF2 (x 1019 n/cm2, (Of)
E > 1.0 MeV)
Surveillance WB1 -U 0.460 0.784 11.88 (0) 9.31 0.614 Weld WB1-W 1.08 1.022 52.14 (30.5) 53.26 1.044 WB1-X 1.75 1.154 45.94 (25.8) 53.00 1.331 WB1 -Z 2.40 1.236 30.23 (13.9) 37.36 1.528 WB2-U 0.614 0.863 49.92 (32.6) 43.10 0.745 Cat. 1 - Z 0.286 0.658 10.99 (1.91) 7.23 0.433 Cat. 1 - Y 1.29 1.071 23.53 (17.79) 25.20 1.147 Cat. 1 - V 2.27 1.222 30.42 (26.50 37.16 1.493 MG2-V 0.302 0.672 45.51 (38.51) 30.58 0.452 MG2-X 1.38 1.089 42.93 (35.93) 46.77 1.187 MG2-U 1.90 1.176 30.81 (23.81) 36.22 1.382 MG2-W 2.82 1.276 50.76 (43.76) 64.76 1.628 SUM:
443.97 12.983 CFwa1d Heat#895075 = L(FF
- ART Nor)+ L(FF2) = (443.97) +(12.983) = 34.2°F Notes:
(a) f = fluence. For Watts Bar Unit 2, see Table 5-6; for Watts Bar Unit 1, see WCAP-18769 (Ref. 13), for Catawba Unit 1, see WCAP-17669 (Ref. 5), and for McGuire Unit 2, see WCAP-17455 (Ref. 14).
(b) FF= fluence factor= f(0-2a-0.,o"log(fJJ (c) The surveillance weld measured ART NDT values have been adjusted for chemistry and irradiation temperature as follows:
Adjusted ART NDT = (ART NDT, Measured + temp. adjustment) X (CF vessel weld +CFsurv. weld),
The temperature adjustments are based on a time-weighted average temperature of the Watts Bar Unit 2 reactor vessel, which is equal to 557°F over the life of the plant. In addition, the potential of a T avg reduction of 7°F will be taken into consideration; thus, a value of 55O°F will be used for the reactor vessel temperature for the adjustments.
For Watts Bar Unit 1, the CF ratio is 1.32 and temp. adjustment is 9°F (559°F-550°F).
For Watts Bar Unit 2, the CF ratio is 1.20 and temp. adjustment is 9°F (559°F-550°F).
For Catawba Unit 1, the CF ratio is 0.79 and temp. adjustment is 12°F (562°F-550°F).
For McGuire Unit 1, the CF ratio is 1.00 and temp. adjustment is 7°F (557°F - 550°F)
Pre-adjusted values are shown in parenthesis.
System REACTOR COOLANT SYSTEM SDD-Nl-68-4001 Description Document Unit 1 / Unit 2 QA Record Appendix B (Page 17 of 21)
Rev.0050 Page 261 of 271 Table 5-6 Fluence Values for the Watts Bar Unit 2 Reactor Vessel Beltline and Extended Beltline Materials 32 EFPY Fluence Material Description (n/cm2, E > 1.0 MeV)
Clad/Base 1/4T Location 3/4T Location Metal Interface (x=2.116 in.)
(x=&.349 in.)
(Inner Surface)
Reactor Vessel Beltline Materials Intermediate Shell Forging 05 1.87E+19 1.13E+19 4.07E+18 Lower Shell Forging 04 1.93E+19 1.16E+19 4.21 E+18 Intermediate to Lower Shell 1.83E+19 1.10E+19 3.99E+18 Circumferential Weld Seam was Reactor Vessel Extended Beltline Materials Upper Shell Forging 06 5.38E+17 3.24E+17 1.17E+17 Bottom Head Ring 03 2.41E+18 1.45E+18 5.25E+17 Upper to Intermediate Shell 6.41 E+17 3.86E+17 5.25E+17 Circumferential Weld Seam was Lower Shell to Bottom Head Ring 2.67E+18 1.61E+18 5.82E+17 Circumferential Weld Seam W04
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix B (Page 18 of 21)
SDD-Nl-68-4001 Rev.0050 Page 262 of 271 Table 5-7 Adjusted Reference Temperature Evaluation for the Watts Bar Unit 2 Reactor Vessel Beltline and Extended Beltline Materials through 32 EFPY at the 1/4T Location<a>
Reactor Vessel CF 1/4Tf 1/4T ARTNDT IRTNDT(b) aib) a"'
M Location (Of)
(n/cm2, FF (Of)
(Of)
(Of)
(Of)
(Of)
E>1.0 MeV)
Reactor Vessel Beltline Materials Intermediate Shell 31.0 1.13E+19 1.033 32.0 14 0.0 16.0 32.0 Forging 05 Lower Shell Forging 31.0 1.16E+19 1.042 32.3 5
0.0 16.2 32.3 04 Intermediate to 54.0 1.10E+19 1.027 55.5
-50 0.0 27.7 55.5 Lower Shell Circumferential Weld SeamW05 Using Surveillance 34.2 1.10E+19 1.027 35.1
-50 0.0 14.0 28.0 Capsule Data<bl Reactor Vessel Extended Beltline Materials Upper Shell Forging 44.0 3.24E+17 0.230 06 10.1
-14 0.0 5.1 10.1 Bottom Head Ring 37.0 1.45E+18 0.495 18.3
-40 0.0 9.2 18.3 03 Upper to 41.0 3.86E+17 0.254 10.4 10 0.0 5.2 10.4 Intermediate Shell Circumferential Weld SeamW06 Lower Shell to 41.0 1.61E+18 0.518 21.3 10 0.0 10.6 21.3 Bottom Head Ring Weld Seam W04 Notes:
(a) Values taken from WCAP-18769-NP (Ref. 13)
(b) The initial RTNorvalues are measured values; therefore, 01= 0°F ART (Of) 78.0 69.6 60.9 13.7 6.2
-3.3 30.8 52.5
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix B (Page 19 of 21)
SDD-Nl-68-4001 Rev.0050 Page 263 of 271 Table 5-8 Adjusted Reference Temperature Evaluation for the Watts Bar Unit 2 Reactor Vessel Beltline and Extended Beltline Materials through 32 EFPY at the 3/4T Location<a>
Reactor Vessel CF 3/4Tf 3/4T ARTNDT IRTNDT(b) aib) a"'
M ART Location (Of)
(n/cm2, FF (Of)
(Of)
(Of)
(Of)
(Of)
(Of)
E>1.0 MeV)
Reactor Vessel Beltline Materials Intermediate Shell 31.0 4.07E+18 0.751 23.3 14 0.0 11.6 23.2 60.5 Forging 05 Lower Shell Forging 31.0 4.21E+18 0.759 23.5 5
0.0 11.8 23.5 52.1 04 Intermediate to 54.0 3.99E+18 0.745 40.2
-50 0.0 20.1 40.2 30.5 Lower Shell Circumferential Weld SeamW05 Using Surveillance 34.8 3.99E+18 0.745 25.5
-50 0.0 12.7 25.5 1.0 Capsule Data<bl Reactor Vessel Extended Beltline Materials Upper Shell Forging 44.0 1.17E+17 0.122 5.4
-14 0.0 2.7 5.4
-3.3 06 Bottom Head Ring 37.0 03 5.25E+17 0.301 11.1
-40 0.0 56 11.1
-17.8 Upper to 41.0 Intermediate Shell 1.40E+17 0.137 5.6 10 0.0 2.8 5.6 21.2 Circumferential Weld SeamW06 Lower Shell to 41.0 Bottom Head Ring 5.82E+17 0.317 13.0 10 0.0 6.5 13.0 36.0 Weld Seam W04 Notes:
(a) Values taken from WCAP-18769-NP (Ref. 13)
(b) The initial RTNoTvalues are measured values; therefore, 01= 0°F
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix B (Page 20 of 21)
SDD-Nl-68-4001 Rev.0050 Page 264 of 271 Table 5-9 Summary of the Limiting ART Values Used in the Generation of the Watts Bar Unit 2 Heatup/Cooldown Curves EFPY Limiting ART(aJ (°F) 1/4T I
3/4T 32 88 I
71 Note:
(a) The ART Values used for heatup and cooldown limit curve development were calculated in WCAP-18191-NP (Ref. 1),
rounded, and increased by 10°F to add additional margin; this approach is conservative. The limiting ART values calculated in WCAP-18769-NP (Ref. 13) were provided in Tables 5-7 and 5-8.
(b) These ART values are bounded by the ART values shown in Table 5-9, which were used to generate the 32 EFPY heatup/cooldown curves.
System Description Document REACTOR COOLANT SYSTEM Unit 1 / Unit 2 QA Record Appendix B (Page 21 of 21)
SDD-Nl-68-4001 Rev.0050 Page 265 of 271 Table 5-10 RT PTS Calculations for the Watts Bar Unit 2 Beltline and Extended Beltline Materials at 32 EFPY Material CF 32 EFPY Fluence ff(a)
IRTNDT ARTNDT au(e) aA(b)
M(cJ RTPTS(d)
(Of)
(n/cm2, (Of)
(Of)
(Of)
(Of)
(Of)
(Of)
E>1.0 MeV)
Reactor Vessel Beltline Materials Intermediate Shell Forging 05 31.0 1.87E+19 1.171 14 36.3 0.0 17.0 34.0 84.3 Lower Shell Forging 04 31.0 1.93E+19 1.180 5
36.6 0.0 17.0 34.0 75.6 Intermediate to Lower Shell 54.0 1.83E+19 1.166
-50 62.9 0.0 28.0 56.0 68.9 Circumferential Weld Seam was Using Surveillance Capsule 34.8 1.83E+19 1.166
-50 39.9 0.0 14.0 28.0 17.9 Data Reactor Vessel Extended Beltline Materials Upper Shell Forging 06 44.0 5.38E+17 0.304
-14 13.4 0.0 6.7 13.4 12.8 Bottom Head Ring 03 37.0 2.41E+18 0.615
-40 22.7 0.0 11.4 22.7 5.5 Upper to Intermediate Shell 41.0 6.41E+17 0.334 10 13.7 0.0 6.8 13.7 37.4 Circumferential Weld Seam was Lower Shell to Bottom Head 41.0 Ring Weld Seam W04 2.67E+18 0.641 10 26.3 0.0 13.1 26.3 62.5 Notes:
(a) FF= fluence factor= f<0-2s-o.1I0g<fl>
(b) Per the guidance of 10 CFR 50.61, the base metal OA = 17°F when surveillance data are non-credible or not used to determine the CF, and the base metal oA = 8.5°F when credible surveillance data are used. Also per 1 0 CFR 50.61, the weld metal OA = 28°F when surveillance data are non-credible or not used to determine the CF, and the weld metal oA =
14°F when credible surveillance data are used. However, OA need not exceed 0.S*ART Nor for either base metals or welds, with or without surveillance data.
(d) M = Margin = 2 * (ou2 + OA2) 1'2 (e) RT PTs = IRT Nor+ ART PTs + Margin (e) As indicated in Table 5-1 of this report, the IRT Nor values are measured; hence, according to 10 CFR 50.61, Ou= 0°F